ML11356A137

New York State (NYS) Revised Pre-Filed Evidentiary Hearing Exhibit NYSR0013E, UFSAR, Rev. 20, Indian Point Unit 3 (Submitted with License Renewal Application) (2007) (IP3 UFSAR, Rev. 20)

ML11356A137
Person / Time
Site:	Indian Point
Issue date:	12/22/2011
From:	State of NY, Office of the Attorney General
To:	Atomic Safety and Licensing Board Panel
SECY RAS
Shared Package
ML11356A134	List: ML11356A135 ML11356A136 ML11356A137 ML11356A138 ML11356A140 ML11356A141 ML11356A142
References
RAS 21609, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01
Download: ML11356A137 (210)

Text

NYSR0013E Revised: December 22, 2011 IP3 FSAR UPDATE In addition, the containment structure will withstand the following Tornado generated missiles (only one missile was considered acting at any time simultaneously with the 360 mph wind load):

Horizontal Missiles

1) 4" x 12' wood plank at 300 mph

2) 4000 Ib auto at 50 mph less than 25' above the ground (25 ft2 contact area).

Vertical Missiles

1) 4" x 12' x 12' wood plank at 90 mph

2) 4000 Ib auto at 17 mph less than 25' above the ground (25 ft2 contact area).

Specific structural effects as the result of missile impact are: 1) missile penetration and 2) structural response to dynamic impact. In addition to the overall structural effects such as overturning moment and base shear, the local structural effects must be considered in the design for tornado wind and generated missile loads. For missile loads, limited local plasticity, structural dynamic response ductility and redistribution of stresses in redundant structures due to plastic action was permitted.

Consideration of tornado loads was not a factor in the design of the Containment structure. The 3 psig negative pressure is approximately 4% of the maximum internal pressure load (1.5P=70.5 psig) thus stresses introduced into the rebar from this load are very small.

5.1.4 Penetrations 5.1.4.1 General In general, a penetration consists of a sleeve embedded in the concrete wall and welded to the containment liner. The weld to the liner is shrouded by a continuously pressurized channel which is used to demonstrate the integrity of the penetration-to-liner weld joint. The pipe, electrical conductor cartridge, duct or equipment access hatch passed through the embedded sleeve and the ends of the resulting annulus were closed off, either by welded end plates, bolted flanges or a combination of these.

Differential expansion between a sleeve and one or more hot pipes passing through it was accommodated by using a bellows type expansion joint between the outer end of the sleeve and the outer end plate, as shown on Figure 5.1-12.

The components are considered ASME Section XI Class MC or CC components and any repair or replacement activities shall be performed in accordance with ASME Section XI Subsections IWE and IWL of the ASME Code, 1992 Edition with certain exceptions whenever specific relief is granted by the NRC.

Pressurizing connections were provided to continuously demonstrate the integrity of the penetration assemblies.

5.1.4.2 20 of 188 IPEC00035441 IPEC00035441

IP3 FSAR UPDATE Electrical Penetrations "Cartridge" type penetrations are used for all electrical conductors passing through the Containment. The penetrations are provided with a pressure connection to allow continuous pressurization. Insulating bushings or fused glass seals are used to provide a pressure barrier for the conductor.

These components are considered ASME Section XI Class MC or CC components and any repair or replacement activities shall be performed in accordance with ASME Section XI Subsections IWE and IWL of the ASME Code, 1992 Edition with certain exceptions whenever specific relief is granted by the NRC.

Figure 5.1-13 shows a design of typical electrical penetrations. There are approximately 60 electrical penetrations.

Piping Penetrations Double barrier piping penetrations are provided for all piping passing through the Containment.

The pipe is centered in the embedded sleeve which is welded to the liner. End plates are welded to the pipe at both ends of the sleeve. Several pipes may pass through the same embedded sleeve to minimize the number of penetrations required. In this case, each pipe is welded to both end plates. A connection to the penetration sleeve is provided to allow continuous pressurization of the compartment formed between the piping and the embedded sleeve. In the case of piping carrying hot fluid, the pipe is insulated and cooling is provided to maintain the concrete temperature adjoining the embedded sleeve at or below 150 F.

These components are considered ASME Section XI Class MC or CC components and any repair or replacement activities shall be performed in accordance with ASME Section XI Subsections IWE and IWL of the ASME Code, 1992 Edition with certain exceptions whenever specific relief is granted by the NRC.

Cooling is provided for most hot penetrations through the use of air-to-air heat exchangers.

These are made in accordance with the ASME UPV Code,Section VIII, by welding together two embossed sheets of 10 gage carbon steel material, the embossments forming coolant passages. The unit is rolled into the form of a cylinder with an outside diameter slightly smaller than the respective inside diameter of the penetration sleeve. The exchanger is placed inside the sleeve and outside the pipe insulation, with the inlet and outlet coolant connections penetrating the sleeve between the outside concrete wall surface and the bellows expansion joint. The coolant to be used is ambient air fed by a centrifugal blower which is backed up with a full sized spare. The isolation features and criteria for piping penetrations are given in Chapter 6. Figure 5.1-12 shows typical hot and cold pipe penetrations.

Loss of cooling for the sleeve is highly improbable. The heat shield has no moving parts, and the cooling air is at low pressure. There are redundant blowers to assure that cooling air is not lost for a significant time. The blowers operate off a diesel bus and can be manually started following a blackout. The thermal insulation on the pipe wall reduces heat flow to the liner sleeve. Operation of the cooling unit can be ascertained by opening the "flow through" connection of the penetration pressurization system on the penetration sleeve and observing the temperature of the cooling air emerging.

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IP3 FSAR UPDATE In order to lose significant structural properties, concrete must be held continuously at 500 to 600 F. The hottest penetrations are the main steam lines, which normally operate at a temperature of 507 F. The results of a two dimensional transient heat transfer analysis indicated that in the improbable case that all cooling air would be lost to the main steam penetrations, the surrounding concrete would reach a maximum temperature of 200 F in approximately 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> and 280 F in approximately 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />. It is highly improbable that cooling air would be lost a very long period of time since the failure of any of the air blower drive motors is alarmed in the control room. Even if the adjoining concrete did reach these temperatures (200 - 300 F), the strength of the structure would not be impaired for two reasons:

1) No credit was taken for the tensile strength of the concrete.

2) These temperatures have substantially no effect on the strength of the penetration sleeve or the reinforcing bar in the area of the penetration.

A total of approximately 80 pipes pass through approximately 50 penetration sleeves, 23 of which are considered thermally hot. In addition, several spare sleeves (capped and pressurized) are provided for the possible future addition of piping.

Equipment and Personnel Access Hatches An Equipment Hatch was provided. It was fabricated from welded steel and furnished with a double-gasketed flange and bolted dished door. The hatch barrel is embedded in the containment wall and welded to the liner. Provision was made to continuously pressurize the space between the double gaskets of the door flanges and the weld seam channels at the liner joint, hatch flanges and dished door. Pressure is relieved from the double gasket spaces prior to opening the joints. The Personnel Hatch is a double door, mechanically-latched, welded steel assembly. A quick acting type, equalizing valve connects the Personnel Hatch with the interior of the containment vessel for the purposes of equalizing pressure in the two systems when entering or leaving the containment. The Personnel Hatch doors are interlocked to prevent both being opened simultaneously and to ensure that one door is completely closed before the opposite door can be opened.

Remote indicating lights and annunciator situated in the Control Room indicate the door position status. An emergency lighting and communication system operating from an external emergency supply is provided in the lock interior. Emergency access to either the inner door from the containment interior or to the outer door from outside, is possible by the use of special door unlatching tools. The design was in accordance with Section VIII of the ASME Code.

These components are considered ASME Section XI Class MC or CC components and any repair or replacement activities shall be performed in accordance with ASME Section XI Subsections IWE and IWL of the ASME Code, 1992 Edition with certain exceptions whenever specific relief is granted by the NRC.

Outage Equipment Hatch (OEH)

Outage Equipment Hatch can be used in place of the Equipment Hatch at Elevation 95'-0" in the Containment Building during outages. The OEH will be attached to the Containment Building using the same attachments for the Equipment Hatch. The OEH door can be closed and sealed in less than 30 minutes and is designed to withstand the radiation release from a fuel handling accident involving recently-irradiated fuel (i.e, fuel subcritical for less than 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />).

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IP3 FSAR UPDATE The OEH is provided with sealed service penetrations for the passage of service lines (i.e.,

compressed air, electricity, fluid carrying hoses, instrumentation, fiber optic cables, etc.).

Special Penetrations

1) Fuel Transfer Penetration A fuel transfer penetration is provided for fuel movement between the refueling transfer canal in the Reactor Containment and the spent fuel pit. The penetration consists of a 20-inch stainless pipe installed inside a 24-inch pipe. The inner pipe acts as the transfer tube. The transfer tube is fitted with a pressurized double gasketed blind flange on the refueling canal end to seal the reactor containment.

The terminus of the tube outside the containment is closed by a standard gate valve.

The outer pipe is welded to the containment liner and provision is made by use of a special seal ring for pressurizing all welds essential to the integrity of the penetration during plant operations. Bellows expansion joints are provided on the pipes to compensate for any differential movement between the two pipes or other structures.

Figure 5.1-14 shows a sketch of the fuel transfer tube.

2) Containment Supply and Exhaust Purge Ducts The ventilation system purge ducts are each equipped with two quick-acting tight-sealing valves (one inside and one outside of the containment) to be used for isolation purposes. The valves are manually opened for containment purging, but are automatically closed upon a signal of high containment pressure or high containment radiation level. The space between the valves is pressurized above calculated peak accident response pressure, while the valves are normally closed during plant operation. See Section 5-3, Containment Ventilation System, and Section 6.4, Containment Air Recirculation Cooling and Filtration System.

These components are considered ASME Section XI Class MC or CC components and any repair or replacement activities shall be performed in accordance with ASME Section XI Subsections IWE and IWL of the ASME Code, 1992 Edition with certain exceptions whenever specific relief is granted by the NRC.

Two solenoid controlled, pneumatically operated butterfly valves are provided for each purge penetration, one on each side of the containment building wall. Two penetrations, one supply and one exhaust, are required. Valves are spring-loaded to fail closed.

The space between the valves is pressurized from the pressurization system through an electrically operated three-way solenoid valve. The pressure is maintained only when valves are closed and must be relieved before butterfly valves can be opened.

Failure to release this pressure will prevent valves from opening.

Failure of any of the valves to open will prevent the fans from running. Tripping or either of the purge fans will automatically close the butterfly valves and pressurize the space between the valves. Failure of any of the valves to close will prevent the adjacent space from being pressurized, and sound the loss-of-pressurization alarm.

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IP3 FSAR UPDATE Loss of pressure for either zone will be displayed by individual indicating lights at the Main Control Board.

The valve control solenoids and pressurization solenoids are controlled from a single control switch on the fan room control panel. The cycle is initiated by setting the control switch to "open" position. This will energize the pressurization alarm.

When the pressure between the valves has been relieved, the valves control solenoids are energized and the valves opened. If for any reason, any of the four valves fail to open within a given time after the cycle is initiated, all four valves will close and pressure will be restored. The circuit is interlocked to prevent inadvertent opening of the valves during S.1. condition.

Once all four valves have been opened, the operator has a pre-determined time (approximately one minute) to start the purge supply fan. Failure to do so will cause all four valves to close.

Position indicating lights for each of the four valves are provided on the Fan Room Control Panel and Main Control Board.

3) Sump Penetrations The piping penetration in the containment sump area is not of the typical sleeve-to-liner design. In this case, the pipe is welded directly to the base liner. The weld to the liner is shrouded by a test channel which is used to demonstrate the integrity of the liner.

5.1.4.3 Design of Penetrations Criteria The liner is basically not a load-carrying member because it is subjected to strains imposed by the reinforced concrete; nevertheless, the liner was reinforced at each penetration in accordance with the ASME Code Section VII. The weldments of liner to penetration sleeve are of sufficient strength to accommodate stress concentrations and adhered strictly to ASM E Code Section VIII requirements for both type and strength.

Liner stress is imposed on the cylindrical penetration as a circular uniform load acting around the circumference of the penetration. The penetration thicknesses were chosen to accommodate this load without causing severe distress at the opening.

The penetration sleeves and plates were designed to accommodate all loads imposed on them under operating conditions (thermal effects and internal penetrations and test pressures) and accident conditions (loads resulting from all strains, internal pressures, and seismic movements).

In the design of the piping penetration sleeves and the piping going through them, maximum total stress in all cases was limited to a value below the yield stress of the material involved; therefore, no plastic design criteria were employed. In particular, piping whose failure would result in a Loss-of-Coolant Accident and the main steam and feedwater pipe penetrations and pipe supports in the Containment Building were designed to prevent the formation of a plastic 24 of 188 IPEC00035445 IPEC00035445

IP3 FSAR UPDATE "hinge" in the pipe should any of these pipes rupture. This was accomplished by effectively anchoring these pipes at 90 elbows connected to all these pipes adjacent to the penetration 0

both inside and outside the building, and by restraining these pipes along their run inside the building and outside the building to the first stop valve. The anchors and restraints were designed to prevent a breach of containment at the piping penetrations should any of these pipes rupture inside, immediately outside, or within the penetration itself. The penetrations were designed to the strength of the pipe and no further considerations are necessary.

To insure that a Loss-of-Coolant Accident acting simultaneously with an earthquake would not result in a breach of containment by causing a failure of one or more pipe penetrations through the Containment Building wall, the following methods were used:

All auxiliary piping attached to the Reactor Coolant System which passes through penetrations in the Containment Building wall must also pass through the circular secondary shield wall approximately fifteen feet inside the building as illustrated in Plant Drawing 9321-F-25012

[Formerly Figure 5.1-2]. The total number of pipes in this category is very limited. They were examined individually and suitable restraints or anchors were used either at or within the secondary shield wall to prevent a Loss-of-Coolant Accident or a failure of one of these pipes within the secondary shield wall from causing the failure of the building penetrations through which the pipes pass. In some cases, it was physically impossible for any conceivable movement of the end of those pipes attached to the Primary Coolant System to be reflected at the building penetration and impose other than ordinary operating loads at these points. In other cases, it was necessary to design restraints for the pipes at the secondary shield wall to withstand the failure of the pipe within the wall in tension. Some auxiliary pipes attached to the Reactor Coolant System are attached at pOints which will not move; for instance, the reactor coolant pump seal water injection pipes and the steam generator blowdown pipes. In general, these have restraints at the secondary shield wall designed for normal loads plus the reaction forces resulting from the double ended rupture of these pipes within the shield wall.

All Containment Building piping penetrations except main steam and feedwater were designed as anchors for the pipes passing through them and transmit piping loads to the reinforced concrete wall. The anchorage strength exceeds the maximum combined forces imposed by the effects on the piping penetration of dead loads, loads induced from a Loss-of-Coolant Accident, thermal expansion of the pipe, penetration air pressure, and earthquake loads.

The piping penetrations were designed to transmit the above combined loadings to the concrete structure without exceeding the yield strength of the penetration steel. Typical penetration details are shown in Figure 5.1-12. Load transfer from the pipe to penetration anchorage is limited to the actual loads induced or to the ultimate strength capacity of the pipe in bending, shear, axial, or torsional loadings.

All piping penetrating the Containment meet the requirements of the USAS B31.1.0 Power Piping Code. In the case of the main steam and feedwater lines, the supports, inside and outside the Containment Buildings to the second isolation valve, were designed so that a failure of anyone of these pipes does not result in breach of containment of the failure of any other main steam or feedwater pipe between the steam generator and the second isolation valve.

The design of all containment building piping penetration sleeves and end plates except the new Steam Generator Blowdown Penetrations (AA, BB, CC, and DD) and Service Water Penetration (SS) was in accordance with the ASME Boiler and Pressure Vessel Code,Section VIII. The Steam Generator Blowdown Penetration Sleeves and end plates and the Service Water 25 of 188 IPEC00035446 IPEC00035446

IP3 FSAR UPDATE Penetration (SS) end plates were designed in accordance with the requirements of the 1986 edition of the ASME Boiler and Pressure Vessel Code,Section III subsection NC.

These components are considered ASME Section XI Class MC or CC components and any repair or replacement activities shall be performed in accordance with ASME Section XI Subsections IWE and IWL of the AMSE Code, 1992 Edition with certain exceptions whenever specific relief is granted by the NRC.

Pipes which penetrate the containment building wall and which are subject to machinery originated vibratory loadings, such as the Reactor Coolant Pumps, had their supports spaced in such a manner that the natural frequency of the piping system immediately adjacent to the penetrations is greater than the dominant frequencies of the pump. Pipe line vibration was checked during preliminary plant operation; and where necessary, vibration dampers were fitted. This checking and fitting effectively eliminates vibrating loads as a design consideration.

Materials The material for penetrations including the Personnel and Equipment Access Hatches, together with the mechanical and electrical penetrations is carbon steel, conforming with the requirements of the ASME Pressure Vessels Code Section VIII, and exhibiting ductility and welding characteristics compatible with the main liner material. The Equipment Hatch, penetration sleeves and Personnel Lock meet the Charpy V-notch impact values for a minimum of 15 ft-Ibs at -50°F.

The stainless steel expansion joints (bellows) of the hot penetration expansion joints were protected from damage in transit and during construction by sheet metal covers fastened in place at the fabricator's shop. These were left in place permanently if there was no interface with nearby piping or equipment.

Due to cracking in the bellows of the Main Steam and Boiler Feedwater penetrations, replacement bellows were installed. The replacement bellows are constructed of improved materials.

The materials making up the penetrations conform to the following specifications:

Minimum Minimum Yield Tensile Strength Strength Item Specification (PSI) (PSI) Elongation

1. Mech. Penetration ASTM A333, 30,000 55,000 35% in 2" Sleeve - 12" Dia. Gr. 1

& under**

2. Mech. - Over ASTM A201 32,000 60,000 22% in 8" 12" Dia.** Gr. B to A300

3. Rolled Shapes+ ASTM A36, 36,000 58,000 20% in 8" ASTM A131 32,000 58,000 21% in 8" Gr. C 26 of 188 IPEC00035447 IPEC0003544 7

IP3 FSAR UPDATE

4. End Plates a) ASTM A300 32,000 60,000 22% in 8" C1.1 FireBox A201, Gr B(1) b) ASTM A240 25,000 70,000 40% in 2" Type 304L+

• • • c) ASTM A516, 32,000 60,000 21% in 8" Gr. 60

5. Fuel Transfer ASTM A240 25,000 70,000 40% in 2" Tube+ Type 304L

6. Bellows+ a) ASTM A312 25,000 70,000 35% in 2" Type 304L b) ASME SB168 35,000 80,000 30% in 2" Inconel 600++

7. Elec. Penetra- ASTM A333 30,000 55,000 35% in 2" tions** Gr. 1

8. Equip. Hatch ASTM A300 32,000 60,000 22% in 8" Insert** C1.1 Firebox A201, Gr. B

• • The Equipment Hatch, penetration sleeves and Personnel Lock were Charpy tested to a minimum of 15 ft-Ibs at -50°F.

+ No specific NDTT requirements

++ Main Stream and Main Feedwater penetrations

• • • Service Water Penetration SS end plates were Charpy V-notch tested to a minimum of 20 ft-Ibs (10f 3 test only) at oaF or lower with a minimum average of three tests of 251bs Minimum Minimum Yield Tensile Strength Strength Item Specification (PSI) (PSI) Elongation

9. Equip. Hatch ASTM A300, 32,000 60,000 22% in 8" Flanges** C1.1 Firebox A201, Gr. B

10. Equip. Hatch ASTM A300 32,000 60,000 22% in 8" Head** Firebox A201, Gr. B 27 of 188 IPEC00035448 IPEC00035448

IP3 FSAR UPDATE

11. Personnel ASTM A300, 32,000 60,000 22% in 8" Hatch** C1.1. Firebox A201, Gr. B

12. Piping Pene- ASTM A442, 32,000 60,000 22% in 8" tration Reinf.* Gr.60

13. Outage Equipment Hatch is designed and built in accordance with ASME Section VIII, 1989 and made of ASTM A516 Grade 60 or higher Grade material. The structural steel members are made of SA 36; Pipe penetrations are made of ASTM A 106 Grade B.

NOTE:

• The liner plates for the shell, bottom and dome were impact tested on a longitudinal section at 15 ft-Ibs at a temperature 30 degrees below the service temperature of -50 F.

D

• • The Equipment Hatch, penetration sleeves and Personnel Lock were Charpy tested to a minimum of 15 ft-Ibs at -50 F.

D Consideration of Jet Loads, Missile Impact and Tornado Loads for Openings The 3'-0" thick crane wall, the 4'-0" and 6'-0" thick Refueling Canal and the 2'-0" thick operating floor are capable of resisting jet force loads and missiles from primary coolant piping. Thus, jet force loads and missiles from the potential failure of the Primary Coolant System are contained within the reactor coolant compartment shield walls and cannot impinge on the containment structure walls; consequently, these loads were not considered in design of large openings. All other missiles terminate inside these concrete shield walls and consequently were not factored into the large opening design. Large openings are shielded or are far enough away to preclude impingement from main steam and feedwater pipe break loads.

Tornado loads are small compared to the seismic loadings. The tornado shear loads from torsion and translational wind force and the overturning moments caused by wind load have a minimum factor of safety of approximately 2.5 when compared with earthquake shears and moments which were used to size the seismic reinforcing bars. The tornado moment and shears are in fact smaller than the minimum earthquake moments and shears considered in design. On this basis, the seismic bars provide more than an adequate mechanism for resisting tornado loads. In addition, tornado loads act independently of other severe loads; therefore, the Equipment Hatch and Personnel Lock reinforced concrete bosses were designed for simultaneous design basis accident and earthquake loads, which were larger than tornado loads, are of more than adequate strength to resist tornado loads.

The containment structure will not be penetrated by the tornado-generated missiles. The concrete sections around large openings are thicker than the 4'-6" Containment wall and so no further consideration of tornado missiles at the large openings was necessary. The large openings have shielded walls of sufficient thickness to protect against tornado missiles.

Consideration of Curvature of the Wail in the Finite Element Analysis Curvature of the containment cylinder wall was included in the finite element analysis for large openings by assigning three coordinates to each node point in the model. This in effect idealizes the structure as a series of chords of a circle with radii equal to the containment cylinder reference surface. Since the widest element in the fine model at the Equipment Hatch 28 of 188 IPEC00035449 IPEC00035449

IP3 FSAR UPDATE opening is 50", or approximately 1% of the total circumference, the chords adequately represent the curvature of the containment surface. Since the shape and stiffness of the structure was accurately represented in the model, all forces and effects were included in the computer output.

The procedures used to design for the six stresses and the justification for all structural elements (rebar) provided to resist the forces or stress resultants outputted by the computer are discussed in detail in Appendix SA. All concrete in tension was considered cracked in the finite element analysis.

5.1.4.4 Leak Testing of Penetration Assemblies A proof test was supplied to each penetration by pressurizing the necessary areas to 54 psig.

This pressure was maintained for a sufficient time to allow soap bubble and Freon sniff tests of all welds and mating surfaces. Any leaks found were repaired and retested; this procedure was repeated until no leak existed.

5.1.4.5 Construction The qualification of welding procedures and welders was in accordance with Section IX, "Welding Qualifications" of the ASME Boiler and Pressure Vessel Code. The repair of defective welds was in accordance with paragraph UW-38 of Section VIII "Unfired Pressure Vessels."

For penetrations between 9" and 18", all the reinforcing bars including primary and secondary vertical bars and diagonal bars are grouped around the penetrations. Due to the continuity of the bars and the relatively small opening size, no special provisions were needed to resist normal, shear and bending stresses. The penetrations are keyed into the concrete, thus creating an edge loading which induces torsion into the walls. The loads are small and the rebar feels little effect from this torsional loading.

For penetrations greater than 18" to 4'-0" the bars are continuous. Since reinforcing is continuous around penetrations, steps were taken to insure that no local crushing of concrete occurred.

From an article, "Detailing and Placing Reinforcing Bars" by Paul F. Rice from Concrete Construction, January 1965, it was determined that in order to prevent local crushing of the concrete a minimum bend diameter of 31 times the bar diameter is required when the reinforcing is stressed to yield. The angle of bend in the rebar determines the force which is transmitted to the concrete in the event the bar tries to straighten out due to tension. For this reason, most bars were bent at 10 degrees except at large penetrations, including the Equipment Hatch, Personnel Lock, main steam and feedwater, and air purge penetrations, where the deviation of the bar from its centerline is too large to permit a 100 bend. In these cases, the bars were bent at 30 degrees but a tie back system was used which prevents a build-up of forces. To further prevent this buildup (in all cases except the equipment hatch penetration) the line of force makes an angle of one-half of the angle of bend, from a horizontal line for the vertical bars and from a vertical line for the horizontal bars and is tangent to the outside of the penetration.

Details of the Personnel and Equipment Hatch design are presented in Section 3.4 of Appendix SA.

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IP3 FSAR UPDATE Concrete was poured in nominal 5' lifts, 360 degrees with no stagger. Approximately one week was allowed to elapse between pours and the surface was left rough, thoroughly cleaned by air blowdown, and all laitance removed. Joints were thoroughly wetted and slushed with a coat of neat cement grout immediately before placing of new concrete except for the exterior of the containment where surfaces were thoroughly wetted but not grouted.

5.1.4.6 Testability of Penetrations and Weld Seams All penetrations, the Personnel Air Lock and the Equipment Hatch were designed with double seals which are normally pressurized at a minimum pressure greater than the calculated peak accident pressure. Individual testing at 115% containment design pressure is also possible.

These components are considered ASME Section XI Class MC or CC components and any repair or replacement activities shall be performed in accordance with ASME Section XI Subsections IWE and IWL of the ASME Code, 1992 Edition with certain exceptions whenever specific relief is granted by the NRC.

The containment ventilation purge ducts are equipped with double isolation valves and the space between the valves is permanently piped up to the penetration pressurization system.

The space can be pressurized to 115% of design pressure when the isolation valves are closed.

The purge valves fail in the closed position upon loss of power (electric or air).

All welded joints in the liner have steel channels welded over them on the inside of the vessel.

During construction, the channel welds were tested by means of pressurizing sections with Freon gas and checking for leaks by means of a Freon sniffer. Most welds are continuously pressurized during power operation at a minimum pressure greater than the calculated peak accident pressure. Liner welds that are not pressurized during power operation are those welds associated with disconnected sections of the Weld Channel Pressurization System. The integrity of the welds associated with any disconnected sections of the Weld Channel Pressurization System is verified by integrated leak rate testing.

Test connections are provided on the Penetration and Weld Channel Pressurization System lines to the Equipment Hatch and Personnel Airlock to allow for leak testing of the PWCP connections.

The use of the weld channel pressurization system may necessitate periodic relief of pressure buildup within the containment, should the system leak into the containment structure.

When pressure relief of the Containment is required during normal operation, it is accomplished using the containment pressure relief line and not the containment purge lines. However, the pressure relief exhaust is routed through charcoal filters which have an iodine removal efficiency of 90.0%. Prior to pressure relief operations, the Containment Auxiliary Charcoal Filter System (see Section 5.3) may be operated to reduce the activity in the containment atmosphere.

Assuming 1% fuel defects and 50 Ibs/day leakage of reactor coolant into the Containment, the containment atmosphere activity has the maximum value of 20.4 x MPC for iodines and 135.5 x MPC for noble gases after approximately 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> of operation of the containment auxiliary charcoal filter system whose efficiency for iodine removal is 90%.

The activity released to the environment as a result of depressurizing the Containment from 1.0 psig to 0 psig at 1500 CFM for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> based on the above abnormal conditions is:

a) For iodines: 8.26 x 10- 15 curies expressed as equivalent 1-131 30 of 188 IPEC00035451 IPEC00035451

IP3 FSAR UPDATE b) For noble gases: 6.68 curies expressed as equivalent Xe-133 The maximum expected operating conditions considered as normal are taken as 0.2% fuel defects and 14.4 gpd leakage of reactor coolant into the Containment. For these conditions, the containment atmosphere activity is 8.16 x MPC for iodines and 65.0 x MPC for noble gases after approximately 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> operation of the containment charcoal filter system whose efficiency for iodine removal is 99%. The activity released to the environment as a result of depressurizing the containment from 1.0 psig to 0 psig at 1500 cfm for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> through the purge line carbon filters (iodine removal efficiency of 99.0%) based on these maximum operating conditions is:

a) For iodines: 2.49 x 10-6 curies expressed as equivalent 1-131 b) For noble gases: 3.21 curies expressed as equivalent Xe-133 5.1.4.7 Accessibility Criteria The Containment is completely closed whenever the core is critical or whenever the primary system temperature is above 200 F, except as required for brief periods necessary to relieve the Containment to keep the pressure below a reasonable level (1-2 psi g) or to purge the Containment in preparation for Containment entry.

Limited access to the Containment through personnel air locks is possible with the reactor at power or with the primary system at hot shutdown for special maintenance or periodic inspections. Access at power would normally be restricted to the areas external to the reactor equipment compartment primarily for inspection and maintenance of the air recirculation equipment, incore instrumentation chamber drives, and instrument calibration.

After shutdown, the Containment vessel is purged to reduce the concentration of radioactive gases and airborne particulates. This purge system was designed to reduce the radioactivity level to doses defined by 10 CFR 20 for a 40-hour occupational work week, within 2-6 hours after plant shutdown. Since negligible fuel defects are expected for this reactor, much less than the 1% fuel rod defects used for design, purging of the Containment is normally accomplished in less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. To assure removal of particulate matter the purge air will be passed through a high efficiency filter before being released to the atmosphere through the purge vent.

The primary reactor shield was designed so that access to the primary equipment is limited by the activity of the primary system equipment and not the reactor.

5.1.5 System Design Evaluation 5.1.5.1 Reliance on Interconnected Systems The containment leakage limiting boundary is provided in the form of a single, carbon steel liner on the vessel having double barrier weld channels and penetrations. Each system whose piping penetrates this boundary was designed to maintain isolation of the Containment from the outside environment. Provisions are made to continuously pressurize penetrations and most weld channels and to monitor leakage from this pressurization.

5.1.5.2 System Integrity and Safety Factors 31 of 188 IPEC00035452 IPEC00035452

IP3 FSAR UPDATE Pipe Rupture - Penetration Integrity The penetrations for the main steam, feedwater, blowdown and sample lines were designed so that the penetration is stronger than the piping system and that the vapor barrier will not be breached due to a hypothesized pipe rupture.

Major Component Support Structures The support structures for the major components were designed to resist all thrust forces, moments and torques associated with either a Reactor Coolant System or main steam pipe break. All primary structural steel elements were designed for stresses not exceeding yield stress due to these forces.

5.1.5.3 Containment Structure Components Analyses The details of radial, longitudinal and horizontal shear analyses for the containment reinforced concrete are given in Section 5.1.3.

5.1.5.4. Performance Capability Margin The containment structure was designed based upon limiting load factors which were used as the ratio by which accident and earthquake loads were multiplied for design purposes to ensure that the loadl'deformation behavior of the structure is one of elastic, low strain behavior. This approach places minimum emphasis on fixed gravity loads and maximum emphasis on accident and earthquake loads. Because of the refinement of the analysis and the restrictions on construction procedures, the load factors primarily provide for a safety margin on the load assumptions. Tabulations of load combinations and load factors utilized in the design which provide an estimate of the margin with respect to all loads are referenced in Section 5.1.2.

5.1.6 Minimum Operating Conditions The minimum operating conditions which are applicable to the Containment System are given in the Technical Specifications.

5.1.7 Containment System Structure-Inspection and Testing Initial Containment Leakage Rate Testing Criterion: Containment shall be designed so that integrated leakage rate testing can be conducted at the peak pressure calculated to result form the design basis accident after completion and installation of all penetrations and the leakage rate shall be measured over a sufficient period of time to verify its conformance with required performance. (GDC 54 of 7/11/67)

After completion of the containment structure and installation of all penetrations and weld channels, integrated leakage rate tests were performed prior to initial plant operations to establish the respective measured leakage rates and to verify that the leakage rate at the peak accident conditions is no greater than 0.075 percent by weight per day of the containment stream-air atmosphere at the calculated peak accident conditions. The leakage rate tests were performed using the absolute method. The duration of each test was not less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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IP3 FSAR UPDATE Periodic Containment Leakage Rate Testing Criterion: The containment shall be designed so that an integrated leakage rate can be periodically determined by test during plant lifetime. (GDC 55 of 7/11/67)

The peak accident pressure integrated leakage rate test is conducted at periodic intervals during the life of the plant, and also as appropriate in the event major maintenance or major plant modifications are made.

A leak rate test at the peak accident pressure using the same test method as the initial leak rate can be performed at any time during the operational life of the plant, provided the plant is not in operation and precautions are taken to protect instruments and equipment from damage.

These components are considered ASME Section XI Class MC or CC components and any repair or replacement activities shall be performed in accordance with ASME Section XI Subsections IWE and IWL of the ASME Code, 1992 Edition with certain exceptions whenever specific relief is granted by the NRC.

Provisions for Testing of Penetrations Criterion: Provisions shall be made to the extent practical for periodically testing penetrations which have resilient seals or expansion bellows to permit leak tightness to be demonstrated at the peak pressure calculated to result from occurrence of the design basis accident. (GDC 56 of 7/11/67)

Penetrations were designed with double seals which are continuously pressurized above accident pressure. The large access openings such as the Equipment Hatch and Personnel Air Lock are equipped with double gasketed doors and flanges with the space between the gaskets connected to the pressurization system. The system utilizes a supply of clean, dry, compressed air which places the penetrations under an internal pressure above the peak calculated accident pressure.

A permanently piped monitoring system is provided to continuously measure leakage from all penetrations.

Leakage from the monitoring system is checked by continuous measurement of the integrated makeup air flow. In the event excessive leakage is discovered, each penetration can then be checked separately at any time.

These components are considered ASME Section XI Class MC or CC components and any repair or replacement activities shall be performed in accordance with ASME Section XI Subsections IWE and IWL of the ASME Code, 1992 Edition with certain exceptions whenever specific relief is granted by the NRC.

Provisions for Testing of Isolation Valves Criterion: Capability shall be provided to the extent practical for testing functional operability of valves and associated apparatus essential to the containment function for establishing that no failure has occurred and for determining that valve leakage does not exceed acceptable limits. (GDC 57 of 7/11/67)

Capability is provided to the extent practical for testing the functional operability of valves and associated apparatus during periods of reactor shutdown.

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IP3 FSAR UPDATE Initiation of containment isolation employs coincidence circuits which allow checking of the operability and calibration of one channel at a time. Removal or bypass of one signal channel places that circuit in the half-tripped mode.

Local leak rate testing of containment isolation valves is performed in accordance with Technical Specification 5.5.15. The Containment Leakage Rate Program is in accordance with the guidance contained on Regulatory Guide 1.163, except as noted in the Technical Specification.

Field and operational inspection and testing were divided into three phases:

1) those taking place during erection of the Containment Building liner; construction tests

2) those taking place after the containment structure was erected and all penetrations were complete and installed; pre-operational tests

3) monitoring during reactor operation; post-operational tests These components are considered ASME Section XI Class MC or CC components and any repair or replacement activities shall be performed in accordance with ASME Section XI Subsections IWE and IWL of the ASME Code, 1992 Edition with certain exceptions whenever specific relief is granted by the NRC.

5.1.7.1 Construction Tests During erection of the liner, the following inspection and tests were performed:

Bottom Liner Plates All liner plate welds were tested for leak tightness by vacuum box. The box was evacuated to at least a 5 psi pressure differential with the atmospheric pressure.

After completion of a successful leak test, the welds were covered by channels. A strength test was performed by applying a 54 psig air pressure to the channels in the zone for a period of 15 minutes.

The zone of channel-covered welds was pressurized to 47 psig with a 20% by weight of Freon-air mixture. The entire run of the channel to plate welds was then traversed with a halogen leak detector.

The sensitivity of the leak detector is 1 x 10-9 standard CC per second. The sniffer was held approximately Y2 inch from the weld and traversed at a rate of about %-inch per second. The detection of any amount of halogen, indicating a leak, required weld repairs and retesting.

After the halogen test was completed all liner welds not accessible for radiography were pressurized with air to 47 psig and soap-tested. Any leaks indicated by bubbles were repaired and retested. Where leaks occurred, welds were removed by arc gouging, grinding, chipping and/or machining, before rewelding. In addition, the zone of channels was held at the 47 psig air pressure for a period of at least two hours. The drop in pressure was not to exceed the equivalent of a leakage of 0.05% of the containment building volume per day. Compensation for change in ambient air temperature was made if necessary.

Vertical Cylindrical Walls and Dome 34 of 188 IPEC00035455 IPEC00035455

IP3 FSAR UPDATE For the liner, a complete radiograph was made of the first 10 feet of full penetration weld made by each welder or welding operation. A minimum of a 12" film "spot" radiograph was made every 50 feet of weld thereafter on the side walls and dome, except where back-up plates are used. The radiograph films were given to United Engineers and Constructors for their review.

When a spot radiograph showed defects that required repair, two adjacent spots were radiographed. If defects requiring repair were shown in either of these, all of the welding performed by the responsible operator or welder was 100% radiographed to determine the end of defect.

The performance and acceptance standards for all radiography is ASME Section VIII, Paragraph UW51.

The liner plate to plate welds were tested for leak tightness by vacuum box techniques. After successful completion of the spot radiography and vacuum box tests and subsequent repair of all defects, the channels were welded in place over all seam welds in a pre-determined zone. A strength test was performed on the liner plate weld and the channel weld by pressurizing the channel with air at 54 psig for 15 minutes. In addition, each zone of channel covered weld was leak tested under the Freon-air mixture at 47 psig.

In location where radiography was not possible, such as the lower courses of shell plates where back-up plates were used, and where liner bottom welds and floor plates were made to angles and tees, the liner fabricator welded on a 2" long overrun coupon. The overrun coupon was chipped off, marked for location and given to United Engineers and Constructors for testing.

These welds are also vacuum box tested.

Welded studs were visually inspected, and at least one at the beginning of each day's work and another at approximately mid-day were bend-tested to 45 degrees for each welder. Studs failing visual or bend-testing were removed.

While the liner is not a pressure vessel, industry experience has shown that leaks in pressure vessels normally occur at jOints. For this reason and following current liner fabrication practice, there was no radiographic or other non-destructive examination of liner plate.

Liner Erection Tolerance Deviations from the allowable erection tolerance standards were located, documented and, in most cases, eliminated during the normal erection of the liner. This was accomplished by jacking against the polar crane wall, utilizing tubular beams, capped by beams of sufficient cross-sectional area to insure against localized buckling of the liner plate. For areas above the concrete polar crane wall, the required tolerances were met and maintained by circular plate wind girders. For the isolated cases where the liner could not be jacked into tolerance, a Non-Conformance Report (NCR) was written and forwarded to the architect-engineer with a complete survey of the area for an engineering evaluation together with a waiver request. This documentation is maintained by the Authority. Only minor deviations were experienced.

Concrete Compression and Slump Testing The compression test samples consisted of six 6" x 12" cylinders for each 100 cu. yd. or portion thereof, per class, per day. A minimum of one set of six cylinders was made for pours of less than 100 cu. yds. Three cylinders were broken at 7 days and 3 cylinders at 28 days. The basis 35 of 188 IPEC00035456 IPEC00035456

IP3 FSAR UPDATE for rejection was failure to develop a minimum compressive strength at 28 days of 15% above the nominal design strength as proven on an average of the three cylinders.

A slump test was performed on each truckload of concrete used in the first four lifts (20 feet) for the containment exterior wall and was recorded for each sample from which compression test cylinders were made. For all other concrete, a slump test was made and recorded for three truckloads of concrete from each class of concrete per 100 cubic yards (or portion thereof) placed per day. A Quality Control inspector was present during the pour and visually checked the concrete from each truck. Any concrete which appeared to be near or over the limit was slump tested. Wet loads were rejected. The maximum slump for all pours was 5 inches except for special pours when specific approval was received from the Architect-Engineer. In no case was the slump permitted to exceed 7 inches.

The statistical results of compression testing for the 28 day breaks were:

a) 100% of the cylinder break tests exceed the minimum requirement b) 75% of the cylinder break tests exceeded the minimum requirement by at least 1000 psi c) 50% of the cylinder break tests exceeded the minimum requirement by at least 1250 psi d) 25% of the cylinder break tests exceeded the minimum requirement by at least 1750 psi e) 10% of the cylinder break tests exceeded the minimum requirement by at least 2250 psi The samples for compression and a slump testing of concrete were taken from the point of discharge from the truck. There was no occurrence of pour removal or concrete rejected from these test results.

Cadweld Splice Test Program In the Cadweld Test Program, tests were performed on production Cadwelds which had been removed (specifically for testing) from the Containment Building after placement. Of the first 141 production Cadwelds tested in this program, all test results were in excess of the minimum specified strengths.

The following test results were obtained from the actual Cadweld test reports submitted to WEDCO from Consolidated Testing Laboratory. Of the Cadwelds tested:

100% had ultimate strengths of at least 79,000 pis 75% had ultimate strengths of at least 95,100 psi 50% had ultimate strengths of at least 97,600 psi 25% had ultimate strengths of at least 102,600 psi 36 of 188 IPEC00035457 IPEC00035457

IP3 FSAR UPDATE 10% had ultimate strengths of at least 105,100 psi A statistical analysis of these results was performed using the methods outlined in Appendix 5A, Section 5.2.1.

The mean value of the ultimate strength of the splices was 99,580 psi with a standard deviation of 9.960 psi and a total range of 32,750 psi. Of the total at least 99% had an ultimate strength of 76,373 psi. No Cadwelds were rejected on the basis of test results from the Cadweld Test Program.

Penetrations Strength and leak tests of individual penetration internals and closures and sleeve weld channels were performed in a similar manner to the above and all leaks repaired and the penetration or weld channel retested until no further leaks were found.

5.1.7.2 Pre-Operational Tests All penetrations, and the welds joining these penetrations to the containment liner and the liner seam welds, were designed to provide a double barrier which can be continuously pressurized at a pressure higher than the calculated peak accident response pressure of the containment.

This blocks potential sources of leakage with a pressurized zone and at the same time provides a means of monitoring the leakage status of the containment which is more sensitive to changes in the leakage characteristics of these potential leakage sources. Certain liner welds are no longer continuously pressurized. Therefore, the leakage status of these welds is no longer continuously monitored. The integrity of these welds is verified by integrated leak rate testing.

After the Containment Building was complete with liner, concrete structures, and all electrical and piping penetrations, Equipment Hatch and Personnel Lock in place, the following tests were performed:

1) Strength Test:

A pressure test was made on the completed building using air at 54 psig. This pressure was maintained on the building for a period of at least one hour. During this test, measurements and observations were made to verify the adequacy of the structural design. For a description of observations, cracks, strain gauges, etc., refer to the Containment Report, Appendix 5A.

2) Integrated Leakage Rate Tests:

Integrated leakage rate tests were performed on the completed building using the absolute method. These leakage tests were performed with the double penetration and weld channel zones open to the containment atmosphere.

3) Sensitive Leak Rate Test:

After it had been assured that there were no defects remaining from construction, a sensitive leak rate test was conducted. The sensitive leak rate test included only the volume of the weld channels and double penetrations. This test is considered more sensitive than the integrated leakage rate test, as the 37 of 188 IPEC00035458 IPEC00035458

IP3 FSAR UPDATE instrumentation used permits a direct measurement of leakage from the pressurized zones. The sensitive leak rate test was conducted with the penetrations and weld channels at a minimum pressure greater than the calculated peak accident pressure and with the Containment Building at atmospheric pressure. The leak rate for the double penetrations and weld channel zones was equal to or less than 0.2% of the containment free volume per day.

In order to verify that the structural response of the Containment to pressure loads is in accordance with design assumptions and to provide assurance that the structure was constructed in accordance with the design to resist pressure loads, a Structural Integrity Test (SIT) was performed.

Readings and measurements were taken at 0 psig, 12 psig, 21 psig, 41 psig and 54 psig (the latter is 115% of the design pressure of 47 psig) during pressurization, and at 41, 18,21,41, and 0 psig during depressurization.

The following gross deformation measurements were taken during the SIT using invar wire extensometers. This provided a means for taking all measurements inside the containment structure thus eliminating effects of weather and temperature. All results were remotely recorded during the test and data was quickly reduced.

a) Radial deformation of the containment wall was measured at 15 locations in the thickened Equipment Hatch boss and the transition area from the thickened boss to the 4'-6" cylinder wall.

b) Diameter change in the containment structure was measured at 10 locations spaced at approximately 10'-0" between elevations 101'-0" and 191'-0".

c) Radial deflection of the containment cylinder wall was measured at elevation 91'-

0".

d) Vertical deflection of the Containment was measured at elevations 95'-0", 143'-0" and 191'-0" and at the apex of the dome. Redundancy was provided for the measurement at the apex of the dome.

Detailed crack measurements were made prior to the test, at peak test pressure of 54 psig, and following depressurization at five areas of the exterior shell, each of at least forty square feet in area. The areas of detailed measurement were: a quadrant of the personnel lock concrete boss, and ten foot wide strips spanning elevations 43'-0" to 48'-0", 115'-0" to 120'-0", and 188'-

0" to 193'-0".

In addition, the exposed surface of the containment shell was visually inspected prior to the test, at 41 psig during the ILRT, and following depressurization. These inspections were for purposes of monitoring the general crack pattern and for specifically following the behavior of the most significant crack.

5.1.7.3 Acceptability of Testing Program AEC Safety Guide No. 18 "Structural Acceptance Test for Concrete Primary Reactor Containments" was followed for testing except in the following areas:

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1) The pattern of measurement points around the largest opening (equipment hatch) were not as shown in Figure C of Safety Guide 18 which indicated 12 points symmetrically located to measure radial and tangential deflections. The Indian Point 3 Structural Integrity Test required taking of radial measurements at 15 locations around the equipment hatch.

Due to access restrictions, no deflection readings were taken on the lower vertical axis of this opening; the 15 measurement locations were symmetrically positioned in the remaining accessible area around this opening. Tangential deflections were not taken, as they were insignificant compared to the radial deflections. The second largest opening (personnel hatch) was structurally loaded in a manner similar to the equipment hatch; no deflection measurements were taken for the personnel hatch opening. This program of radial deflection measurements provided the necessary data to verify that anticipated deformations were taken into account and were within acceptable limits.

2) The structural integrity of the OEH was tested in the Vendor's shop to 7.5 psig for 10 minutes, then the pressure was dropped to 6 psig and the air supply was closed. All tests were performed in accordance with the requirements of ASME B&VP code Section VIII, 1989 Code Part UG-99 or UG-100 for the fabricated Carbon Steel.

5.1.7.4 Post-Operational Tests The double penetrations and most weld seam channels which were installed on the inside of the liner in the Containment are continuously pressurized to provide a continuous, sensitive and accurate means of monitoring their status with respect to leakage. Certain liner welds are no longer continuously pressurized. Therefore, the leakage status of these welds is no longer continuously monitored. The integrity of these welds is verified by integrated leak rate testing.

No periodic structural integrity tests of the Containment are planned. Periodic peak pressure containment integrated leakage rate test (ILRTs) are performed in accordance with the Technical Specifications. Peak pressure tests are to be conducted as appropriate in the event major maintenance or major plant modifications are made. As a prerequisite to the ILRT, a detailed visual examination of the accessible interior and exterior surfaces of the containment structure and its components is required to uncover any evidence of deterioration which may affect the containment integrity. However, no degradation of structural integrity is expected.

The Authority does not consider periodic structural integrity tests as warranted either separately or in conjunction with other tests.

The Containment Leakage Rate Testing Program details requirements for inspection of the accessible interior and exterior surfaces of the containment structure and its components. This periodic surveillance of the Containment and associated structures is visual and includes critical areas as well as a general examination of the accessible surfaces for deterioration. The inspection is also performed prior to any integrated leak test. The insulation attached to the steel liner is designed so that sections can be removed to facilitate inspection of the liner.

These components are considered ASME Section XI Class MC or CC components and any repair or replacement activities shall be performed in accordance with ASME Section XI Subsections IWE and IWL of the ASME Code, 1992 Edition with certain exceptions whenever specific relief is granted by the NRC.

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IP3 FSAR UPDATE Provisions have been made for access to the upper external parts of the containment structure.

These provisions consider the use of movable scaffolding while performing periodic inspection and testing during the service life of the facility.

References

1. Stellmeyer, J. B., W. H. Munse and E. A. Selby, "Fatigue Tests of Plates and Beams with Stud Shear Connections." Highway Research Record, Number 76.

2. Singleton, Robert C. "The Growth of Stud Welding." Welding Engineer, July 1963.

3. United States Atomic Energy Commission - Nuclear Reactors and Earthquakes, TID-7024,1963.

4. Blume, J., N. Newark, L. Corning - Design of Multistory Reinforced Concrete Building for Earthquake Motions - Portland Cement Association.

5. Timoshenko, S., and S. Woinowsky-Kreiger, Theory of Plates and Shells, Second Edition, McGraw-Hili, 1954.

6. American Concrete Institute, Code for Reinforced Concrete Chimney Design, ACI-505.

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IP3 FSAR UPDATE TABLE 5.1-1 FLOODED WEIGHTS - CONTAINMENT BUILDING Item Flooded/Equipment Weight, Ib Pressurizer -1 346,000 Steam Generators - 4 3,816,400 Reactor - 1 (a) Vessel 868,000 (b) Internals 420,000 (c) Piping 1,000,000 Reactor Pumps - 4 824,000 Accumulator Tanks - 4 529,000 175 Ton Polar Crane - 1 650,000 Ventilation Fans - 5 656,000 Reactor Coolant Drain Tank - 1 20,000 Pressure Relief Tank - 1 129,000 Other Miscellaneous Equipment 100,000 TOTAL 9,358,400 41 of 188 IPEC00035462 IPEC00035462

IP3 FSAR UPDATE 5.2 CONTAINMENT ISOLATION SYSTEM 5.2.1 Design Basis Each system whose piping penetrates the Containment's leakage limiting boundary was designed to establish or maintain isolation of the Containment from the outside environment under the following postulated conditions:

a) Any accident for which isolation is required (severely faulted conditions) coincident with b) An independent single failure or malfunction (expected faulted condition) occurring in any active system component within the isolated bounds.

Piping penetrating the Containment was designed for pressures at least equal to the containment design pressure. Containment isolation valves were provided, as necessary, in lines penetrating the Containment to assure that no unrestricted release of radioactivity can occur. Such releases might be due to rupture of a line within the Containment concurrent with a Loss-of-Coolant Accident, or due to rupture of a line outside the Containment, which connects to a source of radioactive fluid within the Containment.

In general, isolation of a line outside the Containment protects against releases due to rupture of the line inside concurrent with a Loss-of-Coolant Accident, and closes off a line which communicates with the containment atmosphere in the event of a Loss-of-Coolant Accident.

Isolation of a line inside the Containment prevents flow from the Reactor Coolant System or any other large source of radioactive fluid in the event that a piping rupture outside the Containment occurs. A piping rupture outside the Containment at the same time as a Loss-of-Coolant Accident is not considered credible, as the penetrating lines are of seismic Class I design up to and including the second isolation barrier and are assumed to be an extension of the Containment.

Normally lines located inside the Containment building that are required to function after an accident are located outside the missile barrier. An exception to this is a portion of the closed loop Component Cooling Water system which is located along the inside of the Crane Wall by the 31 Steam Generator. This is acceptable based on the "Modification of the General Design Criteria 4 requirements for protection against Dynamic effects of postulated pipe ruptures." This takes into account the Leak Before Break methodology, which relaxes the pipe rupture requirements for the Reactor Coolant Loop. The Component Cooling Water piping is also protected by concrete walls from the Pressurizer Surge line located on the other side of the Containment therefore the piping meets the intent of the original design criteria of being protected from credible missiles.

The isolation valve arrangement provides two barriers between the Reactor Coolant System or containment atmosphere, and the environment.

System design is such that failure of one valve to close will not prevent isolation.

The containment isolation valves were examined to assure that they are capable of withstanding the maximum potential seismic loads.

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IP3 FSAR UPDATE To assure their adequacy in this respect:

a) Valves were located in such a manner as to reduce the accelerations on the valves.

Valves suspended on piping spans were reviewed for adequacy for the loads to which the span would be subjected. Valves were mounted in the position recommended by the manufacturer.

b) Valve yokes were reviewed for adequacy, and strengthened as required for the response of the valve operator to seismic loads.

c) Where valves are required to operate during seismic loading, the operator forces were reviewed to assure that system function is preserved. Seismic forces on the operating parts of the valve are small compared to the other forces present.

d) Control wires and piping to the valve operators were designed and installed to assure that the flexure of the line does not endanger the control system.

Appendages to the valve, such as position indicators and operators, were checked for structural adequacy.

e) The design of control systems for automatic containment isolation valves is such that resetting the isolation signal will not result in the automatic reopening of containment isolation valves. Reopening of containment isolation valves requires deliberate operator action.

Containment Isolation Valves Criteria Isolation valves were provided as necessary for all fluid system lines penetrating the Containment to assure at least two barriers for redundancy against leakage of radioactive fluids to the environment in the event of a Loss-of-Coolant Accident. These barriers, in the form of isolation valves or closed systems, are defined on an individual line basis. In addition to satisfying containment isolation criteria, the valving was designed to facilitate normal operation and maintenance of the systems and to ensure reliable operation of other engineered safeguards systems.

Valves utilized in systems for containment isolation service were selected based on tight shutoff requirements, speed of operations, and materials suitable for service in a particular environment relative to temperature, pressure and radiation activity.

The criteria for level of reliability for control valves listed were based on satisfactory operation of the containment isolation valves for the operating life of the plant with the required leak tightness assured by testing and corrective maintenance as required.

The criteria of reliability for swing stop, check, and gate valves listed were based on documented material, quality assurance, compliance for inspection, welding qualification, seismic criteria, testing, and technical specifications for required leak tightness complying with valve manufacturer's standard practice.

Table 5.2.1 provides a summary of containment isolation valve type, actuator, and closure time established for systems penetrating containment.

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IP3 FSAR UPDATE With respect to numbers and locations of isolation valves, the criteria applied were generally those outlined by the seven classes described in Section 5.2.2. Specific containment isolation valves are listed in FSAR Table 5.2-3.

5.2.2 System Design The seven classes listed below are general categories into which line penetrating containment may be classified. The seal water referred to in the listing of categories is provided by the Isolation Valve Seal Water System described in Section 6.5. The following notes apply to these classifications:

1) The "not missile protected" designation refers to lines that are not protected throughout their length inside containment against missiles generated as the result of a Loss-of-Coolant-Accident. These lines, therefore, are not assumed invulnerable to rupture as a result of a Loss-of-Coolant Accident.

2) In order to qualify for containment isolation, valves inside the Containment must be located behind the missile barrier for protection against loss of function following an accident.

3) Manual isolation valves that are locked closed or otherwise closed and under administrative control during power operation qualify as automatic trip valves.

4) A check valve qualifies as an automatic trip valve in certain incoming lines not requiring seal water injection.

5) The double disk type of gate valve was used to isolate certain lines. When sealed by water or gas injection, this valve provides two barriers against leakage of radioactive liquids or containment atmosphere. In certain cases, a double disc valve was used in place of two valves in series having seal water or gas injection between them.

6) In lines isolated by globe valves in series (inboard and outboard) outside containment and provided with seal water injection, the following applies:

a) On process lines ingressing containment (incoming lines) IVSWS will be required to wet the stem packings on both the inboard and outboard valve. IVSW wets the valve plug as well as the stem packing of the RCP seal water injection line containment isolation valves (CH-MOV-250A through D),

b) On process lines egressing containment (outgoing lines) IVSWS will be required to wet only the stem packing on the inboard valve. One exception would be the Steam Generator Slowdown CIVs where both the inboard and outboard valves stem packings are wetted by IVSWS.

7) Excessive loss of seal water through an isolation valve that fails to close on signal is prevented by the high resistance of the seal water injection line. A water seal at the failed valve was assured by proper slope of the protected line, or a loop seal, or by additional valves on the side of the isolation valves away from the Containment.

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8) Lines penetrating containment were designed to the same seismic criteria as the containment vessel up to and including the second isolation barrier. These portions of the penetrating lines are therefore to be considered extensions of the containment.

A review of the Containment Isolation System (NUREG-0578) indicated that there were a number of valves, which automatically reset to the previous position upon reset of containment Phase A isolation. These valves were under operator control via operating procedures to be placed in the closed position prior to resetting of Phase A. Circuits for these valves have been modified to preclude automatic opening on reset. The modification to the valve circuits entailed the installation of pushbuttons that work in conjunction with the containment isolation reset switches so that each valve control circuit has to be reset or the valve will be inhibited from opening.

Class 1 (Outgoing Lines, Reactor Coolant System)

Outgoing lines connected to the Reactor Coolant System which are normally or intermittently open during reactor operation were provided with at least two automatic trip valves in series located outside the Containment. Automatic seal water injection was provided for line in this classification.

Class 2 (OutgOing Lines)

Outgoing lines not connected to the Reactor Coolant System which are normally or intermittently open during reactor operation, and not missile protected or which can otherwise communicate with the containment atmosphere following an accident, were provided, as a minimum, with two automatic trip valves in series outside containment. Automatic seal water injection was provided for lines in this classification with the exception of the reactor coolant pump seal water return line, which was provided with manual seal water injection. Most of these lines are not vital to plant operation following an accident.

Class 3 (Incoming Lines)

Incoming lines connected to open systems outside containment, and not missile protected or which can otherwise communicate with the containment atmosphere following an accident were provided with one of the following arrangements outside containment:

1) Two automatic trip valves in series, with automatic seal water injection. This arrangement was provided for lines, which are not necessary to plant operation after an accident.

2) Two manual isolation valves in series, with manual seal water injection. This arrangement was provided for lines, which remain in service for a time, or are used periodically, subsequent to an accident.

Incoming lines connected to closed systems outside containment, and not missile protected or which can otherwise communicate with the containment atmosphere following an accident were provided either with two isolation valves in series outside containment with seal water injection between them or, at a minimum with one check valve or normally closed isolation valve located either inside or outside containment.

45 of 188 IPEC00035466 IPEC00035466

IP3 FSAR UPDATE The closed piping system outside containment provides the necessary isolation redundancy for lines, which contain only one isolation valve.

Exceptions are the containment spray headers and the safety injection header associated with the boron injection tank, which was valved in accordance with safeguards requirements. The containment spray headers have locked-open double disk gate valves while the safety injection header has either single normally-open double disk gate valves or two normally open gate valves arranged in series.

Class 4 (Missile Protected)

Incoming and outgoing lines which penetrate the Containment and which are normally or intermittently open during reactor operation and are connected to closed systems inside the Containment and protected for missiles throughout their length were provided with at least one isolation valve located outside the Containment. Seal water injection was provided for certain lines in this classification.

Class 5 (Normally Closed Lines Penetrating the Containment)

Lines which penetrate the Containment and which can be opened to the containment atmosphere but which are normally closed during reactor operation were provided with two isolation valves in series or one isolation valve and one blind flange.

Class 6 (Special Service)

There are a number of special groups of penetrating lines and containment access openings.

Some of these are discussed below.

Each ventilation purge duct penetration was provided with two tight-closing butterfly valves, which are closed during reactor power operation and are actuated to the closed position automatically upon a containment isolation or a containment high radiation signal.

One valve is located inside and one valve is located outside the Containment at each penetration. The space between valves is pressurized by air from the Penetration and Weld Channel Pressurization System, whenever they are closed.

The containment pressure relief line is similarly protected. However, since the line can be opened during reactor power operation, three tight closing butterfly valves in series are provided, one inside and two outside the Containment. These valves also are actuated to the closed position upon a containment isolation or containment high radiation signal. The two intravalve spaces are pressurized by air from the Penetration and Weld Channel Pressurization System whenever they are closed.

The equipment access closure is a bolted, gasketed closure, which is sealed during reactor operation. The personnel air locks consist of two doors in series with mechanical interlocks to assure that one door is closed at all times. Each air lock door and the equipment closure were provided with double gaskets to permit pressurization between the gaskets by the Penetration and Weld Channel Pressurization System, Section 6.6.

The fuel transfer tube penetration inside the Containment was designed to present a missile protected and pressurized double barrier between the containment atmosphere and the 46 of 188 IPEC00035467 IPEC00035467

IP3 FSAR UPDATE atmosphere outside the Containment. The penetration closure was treated in a manner similar to the equipment access hatch. A positive pressure is maintained between the double gaskets to complete the double barrier between the containment atmosphere and the inside of the fuel transfer tube. The interior of the fuel transfer tube is not pressurized. Seal water injection is not required for this penetration.

The following lines would be subjected to pressure in excess of the Isolation Valve Seal Water System design pressure (150 psig) in the event of an accident, due to operation of the recirculation pumps:

1) Residual heat removal loop return line

2) Bypass line from residual heat exchanger outlet to safety injection pumps suction

3) Residual heat removal loop sample line

4) Recirculation pump discharge sample line

5) Residual heat removal pump miniflow line

6) Residual heat removal loop outlet line Lines 1, 2, and 6 are isolated by double disc gate valves, while line 3, 4 and 5 are each isolated by two valves in series. These valves can be sealed by nitrogen gas from the high pressure nitrogen supply of the Isolation Valve Seal Water System.

A self contained pressure regulator operates to maintain the nitrogen injection pressure slightly higher than the maximum expected line pressure. The nitrogen gas injection is manually initiated.

Lines which communicate with the containment atmosphere at all times (normally filled with air or vapor) include:

1) Steam jet air ejector return line to containment

2) Containment radiation monitor inlet and outlet lines.

In an accident condition, the space between the two containment isolation valves in each line is sealed by pressurizing with air from the Penetration and Weld Channel Pressurization System.

The air is introduced into each space above the containment calculated peak accident response pressure through a separate line from the Penetration and Weld Channel Pressurization System. Parallel (redundant) fail open valves in each injection line open on the appropriate containment isolation signal to provide a reliable supply of pressurizing air. A flow limiting orifice in each injection line prevents excessive air consumption if one of these valves spuriously fails open, or if one of the containment isolation valves fails to respond to the "trip" signal.

Class 7 (Steam and Feedwater Lines)

These lines and the shell side of the steam generator are considered basically as an extension of the containment boundary and as such must not be damaged as a consequence of Reactor 47 of 188 IPEC00035468 IPEC00035468

IP3 FSAR UPDATE Coolant System damage. This required that the steam generator shell, feed and steam lines within the Containment be classified and designed for the Reactor Coolant System missile-protected category. The reverse is also true in that a steam line break is not to cause damage to the Reactor Coolant System.

5.2.2.1 Isolation Valves and Instrumentation Diagrams Plant Drawings 9321-F-27473, -27203, -27353, -27453, -27503, -27513 Sh, 1, -27363, -27193,

-27233, -27473, -20253, -27263, -70453, -20173, -20193, -27293 Sh. 1 & 2, -27223, -20353, -

40223, -26533, -20363, -26533, and -27243 [Formerly Figures 5.2-1 through 5.2.28] show all valves in lines leading to the atmosphere or to closed systems on both sides of the containment barrier, valve actuation and preferential failure modes, the application of "trip" (containment isolation) signals, relative location of the valves with respect to missile barriers, and the boundaries of seismic Class I designed lines. Figure 5.2-29 defines the nomenclature and symbols used. Individual containment isolation valves are listed in Section 5.2 of the FSAR and Table 5.2-3.

5.2.2.2 Normally Closed Isolation Valves Table 5.2-3 identifies those isolation valves which are either locked closed, or normally closed, (under administrative control) in normal position and relates to Figures 5.2-1 through 5.2-29.

5.2.2.3 Valve Parameters Tabulation A summary of the fluid systems lines penetrating containment and the valves and closed systems employed for containment isolation is presented in Table 5.2-3. Each valve is described as to type, operator, position indication and open or closed status during normal operation, shutdown and accident conditions. Information is also presented on valve preferential failure mode, automatic trip by the containment isolation Signal, and the fluid carried by the line.

Containment isolation valves were provided with actuation and control equipment appropriate to the valve type. For example, air operated globe and diaphragm (Saunders Patent) valves are generally equipped with air diaphragm operators, with fail-safe operation provided by the control devices in the instrument air supply to the valve. Motor operated gate valves are capable of being supplied from reliable onsite emergency power as well as their normal power source.

Manual and check valves, of course, do not require actuation or control systems.

The automatically tripped isolation valves are actuated to the closed position by one of two separate containment isolation signals. The first of these signals is derived in conjunction with automatic safety injection actuation, and trips the majority of the automatic isolation valves.

These are valves in the so-called "non-essential"** process lines penetrating the containment.

This is defined as a "Phase A" isolation and the trip valves are designated by the letter "T" in the isolation diagrams, Figures 5.2-1 through 5.2-29. This signal also initiates automatic seal water injection (See Section 6.5). The second, or "Phase B," containment isolation signal is derived upon actuation of the Containment Spray System, and trips the automatic isolation valves in the so called "essential"* process lines penetrating the containment. These trip valves are designated by the letter "P" in the isolation diagrams.

NOTE:

48 of 188 IPEC00035469 IPEC00035469

IP3 FSAR UPDATE

• "Essential" are those lines required to mitigate an accident, or which, if unavailable, could increase the magnitude of the event. Also, those lines which, if available, would be used in the short term (24 to 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />) to restore the plant to normal operation following an event which has resulted in containment isolation.

• • "Non-Essential" are those lines which are not required to mitigate or limit an accident, which if required at all would be required for long-term recovery only, i.e., days or weeks following an accident.

A manual containment isolation signal can be generated from the Control Room. This signal performs the same functions as the automatically derived "T" signal, i.e., "Phase A" isolation and automatic seal water injection.

Non-automatic isolation valves, i.e., remote stop valves and manual valves, are used in lines which must remain in service, at least for a time, following an accident. These are closed manually if and when the lines are taken out of service.

Standard closing times available with commercial valve models are adequate for the sizes of containment isolation valves used. Valves equipped with air-diaphragm operators generally close in approximately two seconds. The typical closing time available for large motor operated gate valves is ten to thirty seconds. These general closure times are shown on Table 5.2-1.

They are not used for determining valve stroke time limits. Specific design assumptions, closure times for design basis accidents, containment response analyses and resulting off-site dose calculations are contained in specific analyses.

The large butterfly valves used to isolate the containment ventilation purge ducts are each equipped with spring-assisted air pistons capable of closing the valve in two seconds. These valves fail to the closed position on loss of control signal. They also fail closed upon loss of instrument air through use of a local air reservoir as an energy source.

5.2.2.4 Valve Operability All containment isolation valves, actuators and controls are located so as to be protected against missiles which could be generated as a result of a Loss-of-Coolant Accident. Only valves so protected are considered to qualify as containment isolation valves.

Only isolation valves located inside containment are subject to the high pressure, high temperature, steam laden atmosphere resulting from an accident. Operability of these valves in the accident environment is ensured by proper design, construction and installation, as reflected by the following considerations:

1) All components in the valve installation, including valve bodies, trim and moving parts, actuators, instrument air and control and power wiring, were constructed of materials sufficiently temperature resistant to be unaffected by the accident environment. Special attention was given to electrical insulation, air operator diaphragms and steam packing material.

2) In addition to normal pressures, the valves were designed to withstand maximum pressure differentials in the reverse direction imposed by the accident conditions. This criterion was particularly applicable to the butterfly type isolation valves used in the containment purge lines.

49 of 188 IPEC00035470 IPEC000354 70

IP3 FSAR UPDATE 5.2.2.5 Valve Position Indication and Monitoring In general, all remote operated valves have visual position indication in the Control Room.

Table 5.2-4 lists the containment isolation valves and the location of each valve's position indicator lights. Two different types of indicating lights are used: 1) red and green lights and 2) white and red monitoring lights. The red position indicating light is on when the valve is fully open and the green position indicating light is on when the valve is fully closed. At all other positions, both the red and green position indicating lights are on. The red monitor light is on when the valve is in its safeguard position and the white monitor light is on when the motive power is available to the valve. For those valves that are normally de-energized, the white monitor light indicates that power is available to the monitor indicating circuit.

Remote operated containment isolation valves, which are under remote manual control and do not receive a signal from the ESF actuation system, were provided with visual indication of position. An audible alarm feature was provided for remote operated safeguards valves under remote manual control for safeguards functions to denote their off-normal positions.

5.2.2.6 Local Leak Rate Testing of Containment Isolation Valves Local leak rate testing of containment isolation valves is performed in accordance with Technical Specification 5.5.15. The Containment Leak Rate Program is in accordance with the guidance contained in Regulatory Guide 1.163, except as noted in the Technical Specification.

Amendment No. 195 to the Technical Specifications relocated information concerning containment isolation valves from the Technical Specifications to the FSAR.

Subsequent to implementation of Option B of 10 CFR 50, Appendix J, a third-party review of NYPA's (Option B) implementation program was completed. That review, confirmed by NYPA Nuclear Safety Evaluation, determined the scope of the Appendix J, "Type C," Local Leak Rate Test Program was greater than required by regulation. Specifically:

1) Leakage testing of SI-MOV-888A and Band SI-MOV-1835A and B is not required. The valves are not required to be LLR tested for purposes of compliance with Appendix J.

These valves do not represent potential primary containment atmospheric leak paths following a single active failure. Since IVSWS nitrogen will only be applied to these valves in the event of a passive failure, but in no case sooner than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post-LOCA, there are no requirements for performance of leak rate tests.

2) Leak rate testing of the remaining valves penetrations served by the high-pressure nitrogen sub-system IVSWS is not required for compliance with Appendix J. Continued testing is required to ensure adequate nitrogen supply to the affected CIVs for the initial twenty-four hour period following a LOCA.

3) Local leak rate (Type C) testing of SI-1814 A, B, and C is not required for compliance with Appendix J. These valves do not represent potential primary containment atmospheric leak paths following a single active failure.

Note: It is understood the body and packing of the SI-1814 A, B, and C valves are an extension of the containment pressure boundary and are exposed to containment pressure during Type A tests. If that pressure boundary is "broken," i.e., to facilitate 50 of 188 IPEC00035471 IPEC00035471

IP3 FSAR UPDATE calibration of the transmitters, then appropriate testing will be performed to confirm the integrity of the pressure boundary.

4) Local leak rate (LLR) testing of AC-741 is not required for compliance with Appendix J.

This valve does not represent a potential primary containment atmospheric leak path following a single active failure.

5) Local leak rate testing of SI-MOV-885 A & B is not required for compliance with Appendix J. These valves do not represent potential primary containment atmospheric leak paths following a single active failure.

6) LLR testing of the SES CIVs is not required for compliance with Appendix J. Continued testing is required to assure the potential for in-leakage of service water into the containment following a postulated breach of the SWS integrity during the long-term post-LOCA recovery phase is within analyzed limits.

5.2.2.7 Containment Isolation During Refueling Outage The Outage Equipment Hatch (OEH) may be used during an outage, when the permanent Equipment Hatch is removed.

The OEH will maintain containment closure during core alterations and during movement of irradiated fuel assemblies within containment building. The OEH may provide penetrations for temporary services and personnel access during an outage. The penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated on at least one side and are subject to the requirements of ITS 3.9.3.

Electric penetrations used will be verified for leakage and integrity of the pressure boundary connection and not its function.

The roll-up door is an alternate device that is capable of rapid closure. It is effectively an airtight, but not pressure-resistant, door that when closed prevents direct communication between the containment atmosphere and the outside atmosphere.

Subsequent to a loss of RHR cooling as defined in ITS 3.9.4 and 3.9.5, the roll-up door provides rapid containment closure until either cooling is restored, or the main equipment hatch (or OEH) may be installed within four hours.

51 of 188 IPEC00035472 IPEC000354 72

IP3 FSAR UPDATE TABLE 5.2-1 CONTAINMENT CONTROL ISOLATION VALVES Valve Type Actuator Closure*

1500# Globe Reverse Diaphragm 6 sec 1500# Globe Motor 10 sec 1500# D.DV Motor 10/30 sec 150# Gate Motor 10 sec 150# Saunders Direct Diaphragm 2 sec 150# Saunders Reverse Diaphragm 2 sec 150# Globe Solenoid 1.5 sec 150# D.D.V. Motor 10 sec 150# Globe Reverse Diaphragm 6 sec 150# Butterfly Air & Spring 2 sec 600# Plug Air Piston 4 sec 150# Butterfly Air Piston 3.5 sec 300# Gate (RHR V 744) Motor 30 sec 150# Gate (Aux Coolant V 769 and V 797) Motor 30 sec

• Note: Closure times listed are general closure times for valve types shown. They do not form the basis of the safety analysis and are not used in determining valve stroke criteria.

52 of 188 IPEC00035473 IPEC000354 73

IP3 FSAR UPDATE TABLE 5.2-2 NORMALLY CLOSED ISOLATION VALVES DELETED 6/99 53 of 188 IPEC00035474 IPEC00035474

IP 3 FSAR UPDATE TABLE 5.2-3 (Sheet 1 of 9)

COl'lTAINMENT PIPING PENETRATIONS AND VALVING

~ ~ .... ~ . . . IU' ...dL ~ "" * * 'U'YY I ....... I .... I .... ~.u.,,"' ."JL"-JL '"' ..... 'U'U' . . . . . U' .. , , ' "

FIGURE SERVICE AND VALVE ID PENET VALVE OPER. PWR. CONT. POSITION FLUID PEl"ETR. NORM. SHUT- POST POST SEALING :vIIN. TEST NO. PENETRATION or CLOSED CLASS TYPE TYPE FAIL ISOL. INDIC. GAS! DESIGN POSITION DOWN ACCID. ACCID. METHOD TEST FLUID SYSTEM (1) POSITION TRIP CONT. WTR (25) POSITION POSITION USAGE PRESS. (16)

RM (psig) 5.2-1 PRESSURIZER RELIEF TA:.JKTOGAS RC-AOV-549 1 GLOBE AIR FC T Yes G H C 0 C 1\0 Water (A)(4) 47 W A'lALYZER RC-AOV-548 GLOBE AIR FC T Yes C 0 C 1\0 Water (A)(4) W Penetration '"Y" 5.2-1 PRESSURIZER RELIEF TANKN,SUPPLY RC-518 3 CHECK No G C 1\0 43 G Penetration '-Y" RC-AOV-550 DIA AIR FC T Yes 0 0 C 1\0 G 5.2-1 PRESSURIZER RELIEF TANK MAKE-UP RC-AOV-552 3 DIA AIR FC T Yes W C C(9) C C 1\0 Water (A)(4) 47 W Penetration '-Y" RC-AOV-519 DIA AIR FC T Yes C(9) C C 1\0 Waler(A)(4) W 52-2 RESIDUAL HEAT REMOVAL RETURN AC-741 6 CHECK No W H 1\0 (5) NlA N/A Penetration "]" AC-MOV-744 DDV MOTOR FAI Yes 0(8) 0 0 Yes Nitro(M)(32) 43 (15) N 5.2-2 RESID. HEAT REMOVAL LOOP TO SI PUMPS SI- MOV-888A 6 DDV MOTOR FAI Yes W H CIS) LC(28) 0 Yes Nitro(M)(31) NlA N/A Penetration --QQ" SI-MOV-888B DDV MOTOR FAI Yes C(8) LC(28) 0 Yes Nitro(M)(3J) N/A CS 5.2-2 RESID. HEAT REMOVAL LOOP TO SAMPLING SP-AOV-958 6 GLOBE AIR FC T Yes C C(12) C(12) Noll 2) NltrO(M)(32) N SYS SP-AOV-9'59 GLORE AIR FC T Yes W H C C(12) c(12) No(12) Nitro(M)(:l2) 50 N Penetration "QQ" SP-990C GLOBE MANUAL No LC(8) C(12) C(12) No(12) Nitro(M)(32) N 5.2-2 RESID. HEAT REMOVAL LOOP TO RHR PUMP AC-MOV-1870 6 GLOBE MOTOR FAI Yes LTh(8) LTh 0 Yes Nitro(M)(32) N MINlrLOW AC-MOV-743 GATE MOTOR rAI Yes W II O(S) 0 0 Yes Nitro(M)(32) 50 N Penetration "QQ" 5.2-2 RESID. HEAT REMOVAL LOOP OUT AC-732 G DDV MANUAL No W H LC(8) 0 C 1\0 Nitto. 50 (IS) N PenetraTIon -K" lM)(32J 5.2-2 CONTAINMENT SUMP RECIRC. LINE Sl- MOV-8S5A DDV(23) MOTOR FAI Yes W H C(8) C C(18) No(IS) (5) NlA N/A Penetration "00" SI-MOV-885B 5 DDV(23) MOTOR FAI Yes C(8) LC(28) C(18) No(8) (5) NlA 5.2-3 LETDOWN LINE CH-AOV-20l GLOBE AIR FC T Yes W H 0 C(9) C 1\0 Water (A)(4) 47 W Penetra1:ton 'X" CH-AOV-202 1 GLOBE AIR FC T Yes 0 C(9) C 1\0 Water (A)(4) W CS 5.2-3 CHARGING LINE CH-MOV-205 3 GATE MOTOR FAI No 0(8) C(9) C 1\0 Water(M)(4) W Penetration '-R" CH-MOV-226 GATE MOTOR FAI No W C 0(8) C(9) C 1\0 Water(M)(~) ~7 W CH-227 GLOBE MANUAL No LC(8) C C 1\0 Water(M)(4) W CS 5.2-4 REACTOR COOLANT CH-MOV-250A GLOBE MOTOR FAI No 0(8) C(9) C(ll) No(ll) Water(M)(4) W PUMP SEAL WATER CH-MOV-250B GLOBE MOTOR FAI No 0(8) C(9) C(11) NoOl) WaterM)(4) W SUPPLY LINES CH-MOV-250C GLOBE MOTOR FAI No 0(8) C(9) C(ll) No(ll) Water(M)(4) W Penetration "Z" CH-MOV-250D 3 GLOBE MOTOR FAI No W C 0(8) C(9) C(ll) No(ll) Watcr(M)(4) 47 W CH-MOV-441 GLOBE MOTOR FAI No 0(8) C(9) CCll) NOll I) Water(M)(4) W CH-MOV-442 GLOBE MOTOR FAI No 0(8) C(9) C(11) NoOl) Water(M)(4) W

"'U CH-MOV-443 GLOBE MOTOR FAI No 0(8) C(9) C(ll) No(ll) Water(M)(4) W m CH-MOV-444 GLOBE MOTOR FAI No 0(8) C(9) C(ll) No(ll) Watcr(M)(4) W

()

"'U o 54 of 188 m o oo o eN o 01 o0) ~

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IP 3 FSAR UPDATE TABLE 5.2-3 (Sheet 2 of 9)

COl'lTAINMENT PIPING PENETRATIONS AND VALVING Numbers shown in brackets 0 refer to footnotes FIGURE ~ERVICEAAD VALVE ID PENET VALVE OPER. PWR. CONT. POSITION FLUID PENETR NORM. SHUT- POST POST SEALING :vIIN. TEST NO. PENETRATIOI\ or CLOSED CLASS TYPE TYPE FAIL ISOL. INDIC. GAS! DESIGN POSITION DOWN ACCID. ACCID. METHOD TEST FLUID SYSTEM (1) POSITION TRIP CONT. WTR. (25) POSITION POSITION USAGE PRESS. (16)

RM (psig) 5.2-4 REACTOR COOLANT PUMP SEAL WATER CH-MOV-222 2 DDV MOTOR FAI P Yes W C 0(11) 0 C(11) Noell) Water (M)(4) 47 W RETURN PenetraTIon *'R" 5.2-5 REACTOR COOLANT SYSTEM SAMPLE LINES SP-AOV-056E 1 GLOBE AIR FC T Yes W H 0 C C No Water (A)(4) 47 W PeneII UliOil "w" SP-AOV-056F GLOBE AIR FC T Yes 0 C C No Waler (A)(4)

2-5 FUEL TRANSFER TUBE 6 BLIND W H (l7)

Penetrallon '-HH" FLA'IGE (27) 5.2-6 CONTAINME'IT SPRAY SI-869A 3 DDV MANUAL No LO(S) C 0 Yes Water(M)(4) 47 W HEADERS SI-869B DDV MANUAL No LO(8) C 0 Yes Water(M)(4) 47 W Penetrations "GG' SI-867A CHECK No W C Yes 43 G and "P" SI-867B CHECK No Yes 43 G SI-878A GLOBE MANUAL No LCCS) C C Yes 43 G SI-878B GLOBE MANUAL No LCCS) C C Yes 43 G 5.2-7 SAFETY INJECTIOK SI-MOV-1835A DDV MOTOR FAI S Yes 0(8) C 0(10) Yes(33) Ki~o.(M)(33) NlA N/A HEADERS SI-MOV-1835B 3 DDV MOTOR FAI S Yes W H 0(8) C 0(19) Yes(33) Krtro.(M)(33) NlA N/A Penetrations "Q" and '-NN" ST-MOV-8jl A nnv MOTOR FAT Yes 0(8) C- 0(19) Yes(lo) Water (M)(4) 47 W SI- MOV-850C GATE MOTOR FAI Yes LO(8) C 0(19) Yes(19) Water (M)(4) 47 W SI-MOV-850A GATE MOTOR FAI Yes LO(8) C 0(10) Yes(l9) Waler(M)(4) 47 W 5.2-7 SAFETY INJECTIOK TEST SI-859A 5 GLOIlE MANUAL No W C LCCS) C C No Water (A)(4) 47 W Penetration .'y" SI-859C GLOBE MANUAL No LCCS) C C No Water (A)(4) 47 W 5.2-8 ACCUMULATOR NITROGEN SUPPLY NNE-IGlO 5 CHECK No G C No 43 G Penetranon "RR' NNE-AOV-863 GLOBE AIR FC T Yes C(9) C C No 43 G 5.2-8 ACCUMULATOR SAMPLE SP-AOV-956G 2 GLOBE AIR FC T Yes W C C(12) C C(12) Noel2) Water (A)(4) 47 W PenetratIon "RR'* SP-AOV-956H GLOBE AIR FC T Yes C(12) C C(12) Noel2) Water (A)(4) 47 W 5.2-9 PRIMARY SYSTEM VENT WD-AOV-1786 2 DIA AIR FC T Yes 0(9) C C No Water (A)(4) 47 W AND NITROGEN SUPPLY WD-AOV-I787 DIA AIR FC T Yes G H 0(9) C C No Water (A)(4) 47 W Penetration -'V" WD-AOV-1610 3 DIA AIR FC T Yes 0 0 C No 43 G WD-1616 CHECK No 43 G 5.2-9 REACTOR COOLANT DRAIN TK. TO GAS WD-AOV-I78S 2 DIA AIR FC T Yes G H C(13) 0 C No Water (A)(4) 47 W ANALYZER WD-AUV-ln9 DlA AIR FC "j' Yes C(13) C(13) C No Water (A)(4) 47 W Penetration -'V" 5.2-9 RCDT PUMP DISCHARGE WD-AOV-1702 2 DIA AIR FC T Yes W C C(9) 0 C No Water (A)(4) 47 W Penetnltion "Z" WD-AOV-1705 DIA AIR FC T Yes C(9) 0 C No W8ter(A)(4) 47 W 52-10 REACTOR COOLANT

"'U m PUMP COOLING WATER IN AC-MOV-797 AC-MOV-769 3 GATE GATE MOTOR MOTOR FAI FAI P

P Yes Yes W C 0(11) 0(11)

CO!)

cell)

CO!)

C(11)

NoO!)

Noell)

Water (Ml(4)

Water (M)(4) 47 47 W

W

() Penetration -'N"

"'U o m o oo o 55 of 188 eN o 01 o0) ~

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IP 3 FSAR UPDATE TABLE 5.2-3 (Sheet 3 of 9)

COl'lTAINMENT PIPING PENETRATIONS AND VALVING Numbers shown in brackets 0 refer to footnotes FIGURE ~ERVICEAAD VALVE ID PENET VALVE OPER. PWR. CONT. POSITION FLUID PENETR NORM. SHUT- POST POST SEALING :vIIN. TEST NO. PENETRATIOI\ or CLOSED CLASS TYPE TYPE FAIL ISOL. INDIC. GAS! DESIGN POSITION DOWN ACCID. ACCID. METHOD TEST FLUID SYSTEM (1) POSITION TRIP CONT. WTR. (25) POSITION POSITION USAGE PRESS. (16)

RM (psig) 52-10 REACTOR COOLANT AC-MOV-784 2 GATE MOTOR FAI P Yes W C 0(11) C(ll) C(1I) Noell) Water (M)(4) 47 W PUMP COOLING WATER AC-MOV-786 GATE MOTOR FAI P Yes 0(11) C(ll) C(1I) Noell) Water (M)(4) 47 W OUT 6" Penetration *'0" 5 2-10 REACTOR COOLANT AC-FCV-625 2 GATE MOTOR FAI P Yes W C 0(11) C(ll) C(II) Noell) Waler(M)(4) 47 W PUMP LUULlNG WAl'EK AC-MUV-7~9 GAl'E MOTUK FAI P Yes 0(11) CCll) CCII) No(1l) Water (M)(4) 47 W OUT 3" Penetration -'0" 5 2-11 RESIDUAL HEAT EXCHANGERS COOLI'IG AC-75IA 4 CHECK No Yes WA NtA WATERlN AC-75IB CHECK No W C Yes WA N!A Penetrations "KK" and CS "VV" 5 2-11 RESIDUAL HEAT EXCHANGERS COOLI'IG AC-MOV-822A GATE MOTOR FAI S Yes W C C(8) 0 0 Yes WA N!A WATER RETURN AC-MOV-822B 4 GATE MOTOR FAI S Yes C(8) 0 0 Yes WA N!A PenetratIOns" JJ" and "UU" CS 52-12 RECIRC PUMP COOLING AC-752F 4 GLOBE MANUAL No C 0(8) 0 0 Yes WA N!A WATER SUPPLY AC-753F GLOBE MANUAL No W 0(8) 0 0 Yes WA N!A Penetration '"T,T ," CS 52-12 RECIRC PUMP COOLING AC-752I 4 GLOBE MANUAL No W C 0(8) 0 0 Yes WA NtA WATER RETURN AC-753! GLOBE MANUAL No 0(8) 0 0 Yes WA N!A Penetration '"LL" CS 52-13 EXCESS LETDo\\,N HEAT EXCHANGER AC-AOV-791 *1 DIA AIR FC T Yes W C C(9) 0 C No Water O\)(~) ~7 W COOLING WATER IN AC-AOV-79S DIA AIR FC T Ye~ C(9) 0 C No Water (.'1.)(4) 47 W PenetraTIon **U" 52-1l EXCESS l.FTDOVm HEAT EXCHANGER AC-AOV-796 4 GLOBE AIR FC T Yes W C C(9) 0 C No Water (A)(4) 47 W COOLING WATER OUT AC-AOV-793 DIA AIR FC T Yes C(9) 0 C No Water (.'1.)(4) 47 W PenetratIon *'R" 5 2-13 CONTAINMENT SUMP PUMP DISCIIARGE WD-AOV-I728 2 DIA AIR FC T Yes W C 0 0 C No Water (.'1.)(4) 47 W PenetratIon .'y" WD-AOV-I723 DIA AIR FC T Yes 0 0 C No Water (A)(4) 47 W 52-H CONTAINMENT AIR SAMPLE IN- RAD. VS-PCV-1234 6 DIA AIR FC T Ye~ G C 0 0 C(20) No(20) Air (A)(7) 43 G MONITORING SYSTEM VS-PCV-1235 DIA AIR FC T Yes 0 0 C(20) No(20) Air (A)(7) 43 G Penetration "RR'*

5 2-14 CONTAINMENT AIR VS-PCV-1236 6 DIA AIR FC T Yes G C 0 0 C(20) No(20) Air (A)(7) 43 G SAMPLE OUT - RAD. VS-PCV-1237 DIA AIR FC T Yes 0 0 C(20) No(20) Air (A)(7) 43 G MONITORING SYSTEM Penetration "RR'*

5 2-14 AIR EJECTOR CA-PCV-I229 6 GLOBE AIR FC T Yes G C C C C No Air (A)(7) 43 G DISCHARGE TO CA-PCV-1230 GLOBE AIR FC T Yes C C C No Air (A)(7) 43 G CONTAINMENT

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IP 3 FSAR UPDATE TABLE 5.2-3 (Sheet 4 of 9)

COl'lTAINMENT PIPING PENETRATIONS AND VALVING Numbers shown in brackets 0 refer to footnotes FIGURE ~ERVICEAAD VALVE ID PENET VALVE OPER. PWR. CONT. POSITION FLUID PENETR NORM. SHUT- POST POST SEALING :vIIN. TEST NO. PENETRATIOI\ or CLOSED CLASS TYPE TYPE FAIL ISOL. INDIC. GAS! DESIGN POSITION DOWN ACCID. ACCID. METHOD TEST FLUID SYSTEM (1) POSITION TRIP CONT. WTR. (25) POSITION POSITION USAGE PRESS. (16)

RM (psig) 52-15 MAIN STEAM HEADERS CS 7 G H (22) Yes(22)

Penetrations "A,B,C and D" MAIN STEAM TO AUX. FW PUMP CS G Yes TURBINE 52-15 MAIN FEEDWATER HEADERS CS 7 W H Yes Penetrations "E,F,G and H" AUXILIARY FW TCRBINE DRIVEN CS W Yes AUXILIARY FW MOTOR DRIVEN CS W Yes 52-15 STEAM GENERATOR BD-PCV-1214 GLOBE AIR FC T Yes 0 C C No Water (A)(4) 47 W BLOWDOWN BD-PCV-1215 2 GLOBE AIR FC T Yes 0 C C No Water (A)(4) 47 W Penetrations "AA,BB, BD-PCV-1216 GLOBE AIR FC T Yes 0 C C No Water (A)(4) 47 W CC,andDD" BD-PCV-1217 GLOBE AIR FC T Yes 0 C C No Water (A)(4) 47 W W H BD-PCV-1214A GLOBE AIR FC T Yes 0 C e No Water (A)(4) 47 W IlD-PCV-12ISA GLOIlE AIR PC T Yes 0 C e No Water (A)(4) 47 W BD-PCV-1216A GLOBE AIR FC T Yes 0 C e No Water (A)(4) 47 W BD-PCV-1217A GLOBE AIR FC T Yes 0 C e No Water (A)(4) 47 W S 2-IS STEAM GENERATUR BU-PCV-1223 2 GLUBE AlR FC T Yes U L C No Water (A)(4) 47 IV BLOWDOWN BD-PCV-I224 GLOBE AIR FC T Yes 0 C e No Water (A)(4) 47 W SAMPLE BD-PCV-122S GLOBE AIR FC T Yes W H 0 C e No Water O\)(~) ~7 W Four Line~ @ BD-PCV-122G GLOBE AIR FC T Ye~ 0 C e No Water (A)(4) 47 W Penetranon "WOO BD-PCV-I223A GLOBE AIR FC T Yes 0 C e No Water (A)(4) 47 W BD-PCV-122~A GLOBE AIR FC T Yes 0 C e No Water(A)(~) ~7 W BD-PCV-122SA GLOBE AIR FC T Ye~ 0 C e No Water (A)(4) 47 W BD-PCV-l226A GLOBE AIR Fe T Yes 0 C e No Water (A)(4) 47 W S 2-16 VENTILATION SYSTEM SWN-41-1 BV MANUAL No 0(8) 0 0 Yes (6) 47 W COOLING WATER IN SWN-41-2 BV MANUAL No 0(8) 0 0 Yes (6) 47 W Penetrations "La,T.h,Lc, SWN-41-:J BV MANUAl. No W C O(S) 0 0 Yes (Ii) 47 W Ldand Le" SWN-41-4 BV MANUAL No 0(8) 0 0 Yes (6) 47 W SWN-41-S BV MANUAL No 0(8) 0 0 Yes (6) 47 W SWN-4:J-l GATE MANUAl. No C(S) (" C No (Ii) 47 W SWN-43-2 GATE MANUAL No C(8) C e No (6) 47 W SWN-43-3 4 GATE MANUAL No C(8) C e No (6) 47 W SWN-43-4 GATE MANUAL No C(8) C C No (6) 47 W SWN-43-5 GATE MANUAL No C(8) C e No (6) 47 W SWN-42-1 RV No (6) 47 W SWN-42-2 RV No (6) 47 W SWN-42-:J RV No (Ii) 47 W SWN-42-4 RV No (6) 47 W

"'U SWN-42-S RV No (6) 47 W m

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IP 3 FSAR UPDATE TABLE 5.2-3 (Sheet 5 of 9)

COl'lTAINMENT PIPING PENETRATIONS AND VALVING Numbers shown in brackets 0 refer to footnotes FIGURE ~ERVICEAAD VALVE ID PENET VALVE OPER. PWR. CONT. POSITION FLUID PENETR NORM. SHUT- POST POST SEALING :vIIN. TEST NO. PENETRATIOI\ or CLOSED CLASS TYPE TYPE FAIL ISOL. INDIC. GAS! DESIGN POSITION DOWN ACCID. ACCID. METHOD TEST FLUID SYSTEM (1) POSITION TRIP CONT. WTR. (25) POSITION POSITION USAGE PRESS. (16)

RM (psig)

S 2-16 VENTILATION SYSTEM SWN-44-1 BV MANUAL No LTh(8) LTh LTh Yes (6) 47 W COOLING WATER OUT SWN-44-2 BV MANUAL No W C LTh(8) LTh LTh Yes (6) 47 W PenetraTIons "Ma, Mb, SWN-44-3 BV MANUAL No LTh(8) LTh LTh Yes (6) 47 W Me. Mel Me. and SS" SWN-44-4 BV MANUAL No LTh(S) LTh LTh Yes (6) 47 W SWN-44-S BV MANUAL No LTh(8) LTh LTh Yes (6) 47 W SWN-Sl-l GATE MANUAL No 0(8) 0 0 Yes (6) 47 W SWN-SI-2 4 GATE MANUAL No O(S) 0 0 Yes (6) 47 W SWN-SI-3 GATE MANUAL No 0(8) 0 0 Yes (6) 47 W SWN-SI-4 GATE MANUAL No 0(8) 0 0 Yes (6) 47 W SWN-Sl-S GATE MANUAL No 0(8) 0 0 Yes (6) 47 W SWN-71-1 GLOBE MANUAL No Th(8) Th Th Yes (6) 47 W SWN-71-2 GLOIlE MANUAL No Th(8) Th Th Yes (6) 47 W SWN-71-3 GLOBE MANUAL No Th(8) Th Th Yes (6) 47 W SWN-71-4 GLOBE MANUAL No Th(S) Th Th Yes (6) 47 W SWN-71-S GLOBE MANUAL No Th(8) Th Th Yes (6) 47 W CS S 2-17 STATION AIR SA-2*1-1 3 DIA MANUAL No G C LC(8) LC(8) LC No Water(A)(~) ~7 W PeneiIalioIl -'Y" SA-24-2 DIA MANUAL No LC(S) LC(S) LC No Wale! (A)(4) 47 W S 2-17 WELD CHANl\EL PS-PCV-llll-l 4 BALL MANUAL No LO(8) LO LO Yes (l7) WA N/A PENETRATION PS-PCV-llll-2 BALL MANUAL No G C LO(8) LO LO Yes (l7) WA N/A PRESSURE SYSTEM CS (mside)

PenetratIon -'Y" CS (outsIde) 52-19 PURGE SUPPLY DUCT VS-FCV-1170 6 BV AIR FC T(2) Yes G C C 0 C No AiI(A)(7) 43 G VENTILATION VS-FCV-1171 BV AIR FC T(2) Yes C 0 C No AiICA)(7) 43 G Penetration '-EE" S 2-19 PURGE EXHAUST DUCT VS-FCV-II72 6 BV AIR FC T(2) Yes G C C 0 C No AiI(A)(7) 43 G VENTILATION VS-FCV-1173 BV AIR FC T(2) Yes C 0 C No Ai! (A)(7) 43 G Fenetra1:ton "FF" S 2-19 CONTAINMENT VS-PCV-1190 BV AIR FC T(2) Yes G C C(14) C C No AiI(A)(7) 43 G PRESSURE RELIEF VS-PCV-1191 6 BV AIR FC T(2) Yes C(14) C C No AiI(A)(7) 43 G VENTILATION VS-PCV-1192 BV AIR FC T(2) Yes C(l4) C C No AiriA)(7) 43 G Penetration "PP" 5 2-20 RECIRCULATION PUMP SP-MOV-990A b GATE MOTOR FAI No W C LC(8) C LC (12) No NitrolM)(32) 50 N DISCHARGE SAMPLE SP-MOV-990B GATE MOTOR FAI No LC(S) C LC(l2) No N!trOlM)(32) 50 N LINE Penetration '"TT" 5 2-20 PRESSURIZER STEAM SP-AOV-956A 1 GLOBE AIR FC T Yes W H C C C No Water (A)(4) 47 W SAMPLE LINE SP-AOV-956B GLOBE AIR FC T Yes C C C No Water (A)(4) 47 W Penetration "W" S 2-20 PRESSURIZER LIQUID SP-AOV-956C 1 GLOIlE AIR FC T Yes W II C C C No Water (A)(4) 47 W

"'U SAMPLE LINE SP-AOV-9S6D GLOBE AIR FC T Yes C C C No Water (A)(4) 47 W m Penetration "W"

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IP 3 FSAR UPDATE TABLE 5.2-3 (Sheet 6 of 9)

COl'lTAINMENT PIPING PENETRATIONS AND VALVING Numbers shown in brackets 0 refer to footnotes FIGURE ~ERVICEAAD VALVE ID PENET VALVE OPER. PWR. CONT. POSITION FLUID PENETR NORM. SHUT- POST POST SEALING :vIIN. TEST NO. PENETRATIOI\ or CLOSED CLASS TYPE TYPE FAIL ISOL. INDIC. GAS! DESIGN POSITION DOWN ACCID. ACCID. METHOD TEST FLUID SYSTEM (1) POSITION TRIP CONT. WTR. (25) POSITION POSITION USAGE PRESS. (16)

RM (psig) 5.2-21 CONTAINMENT PRESSURE SI-1814A 6 GLOBE MANUAL No G C LO(8) 0 0 Yes (34) NIA INSTRUMENTATION CS LINE Penetration "RR'*

5 2-21 CONTAINMENT PRESSURE SI-1814B 6 GLOBE MANUAL No G C LO(8) 0 0 Yes (34) NIA INSTKUMEN'I'Al'lUN CS LINE Penetration'LL" 5 2-21 CONTAINMENT PRESSURE SI-ISI4C 6 GLOBE MANUAL No G C LO(S) 0 0 Yes (34) NtA INSTRUMENTATION CS LINE PenetratIon -'0" 52-22 POST ACCIDENT SP-SOV-S06 GLOBE SOL. FC T(IO) Yes C(S) C C(I2) Yes (I 2) AiriA)(7) 43 G CONTAINMENT SP-SOV-SOl GLOBE SOL. FC TOO) Yes G C C(8) C C(l2) Yes (12) Air (A)(7) 43 G SAMPLING SCPPLY A'ID SP-SOV-SOS GLOBE SOL. FC TOO) Yes C(8) C C(l2) Yes (12) Air (A)(7) 43 G RETURN LINES SP-SOV-512 GLOBE SOL. FC TOO) Yes e(8) C eel 2) Yes (12) Air (A)(7) 43 G Penetrations "R, TT, SP-SOV-513 GLOBE SOL. FC T(IO) Yes C(S) C C(I2) Yes (I 2) AiriA)(7) 43 G LL, Z, and 0" SP-SOV-SII 5 GLOBE SOL. FC TOO) Yes C(8) C C(l2) Yes (12) Air (A)(7) 43 G SP-SOV-SI6 GLOBE SOL. FC TOO) Yes C(8) C C(l2) Yes (12) Air (A)(7) 43 G SP-SOV-S09 GLOBE SOL. FC mO) Yes C(S) C C(l2) Yes (12) Air (A)(7) 43 G SP-SOV-5IO GLOBE SOL. FC TOO) Yes C(8) C C(l2) Yes (12) Air (A)(7) ~3 G SP-SOV-SI4 GLOBE SOL. FC TOO) Yes e(8) C eel 2) Yes (12) AirCA)(7) 43 G SP-SOV-SIS GLOBE SOL. FC TOO) Yes e(8) C eel 2) Yes (12) Air (A)(7) 43 G CONTAINMENT SUMP RECIRCULATION CS (3) 6 C (17)

( SPARE)

Penetr8tion "0 1 OJ

52-25 INSTRUMENT AIR P.AVENTING IA-39 6 CHECK No G C No 43 G SYSTEM SUPPLY IA-PCV-1228 DIA AIR FC T Yes 0 0 C (24) No (24) 43 G Penetration -'Y" 5 2-25 POST ACCIDENT PS-7 DIA MANUAL No Le(8) LC C (24) No (24) (17) 43 G VENTING SYSTEM 5 G C EXHACST LINE PS-8 DIA MANUAL No LC(S) LC C (24) No (24) (I 7) 43 G PenetratIOn 'LL" PS-O DIA MANUAL No Le(8) LC C (24) No (24) (17) 43 G PS-IO DIA MANUAL No LeeS) LC C (24) No (24) (17) 43 G 52-26 CONTAINMENT LEAK TEST INSTRUMENT r:S (3) G r: No (17)

SENSOR LINE TlueeliIles@

PenetratIon "RR'*

5 2-26 CONTAINMENT LEAK TEST AIR LINE CS (3) 6 G C (30) No (17)

PenetraTIons -'XX and YY" CB-7 BALL MANUAL C(S) C C No (I 7) 43 G 52-27 EQUIPMENT CB-8 BALL MANUAL (29) G C C(8) C C No (17) 43 G ACCESS CB-5 6 CHECK(26) - 43 G CIl-6 CHELK(26) 43 G CB-3 BALL MANUAL C(8) C C No (17) 43 G 5.2-27A PERSONNEL CB-4 BALL MANUAL (29) G C C(8) C C No (17) 43 G AIRLOCK CB-I 6 CHECK(26) 43 G

"'U CB-2 CHECK(26) m 52-28 DE'vIIN. WTR INTO DW-AOV-I PLUG AIR FC T Yes C C(21) C No Water (A)(4) 43 47 G

W

() CONTAINMENT DW-AOV-2 6 PLUG AIR FC T Yes W C C C(21) C No Water (A)(4) 47 W

"'U o Penetration .'y" m o 59 of 188 oo o eN o 01 o0) ~

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IP 3 FSAR UPDATE TABLE 5.2-3 (Sheet 7 of 9)

COl'lTAINMENT PIPING PENETRATIONS AND VALVING Numbers shown in brackets 0 refer to footnotes ABBREVIATIONS:

A Automatic AMB Ambient BV Butterfly Valve C Cold CS Closed System COL Check Off List DDV Double Disc Gate Valve DIA Diaphragm Valve FAI Fail As Is FC Fail Closed FO Fail Open G Gas H Hot LC Locked Closed LO Locked Open LTh Locked Throttled M Manual N Nitrogen POP Plant Operating Procedures P Containment Isolation Signal Phase B T Containment Isolation Signal Phase A Th Throttled RV Relief Valve S Safety h*ction Signal (Opens valves on SI signal)

SOP System Operating Procedures SOL Solenoid Operated Valves W Water 60 of 188

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IP 3 FSAR UPDATE TABLE 5.2-3 (Sheet 8 of 9)

COl'lTAINMENT PIPING PENETRATIONS AND VALVING Numbers shown in brackets 0 refer to footnotes DEFINITIONS:

NORMAL POSITION: Defined as RCS operation above 200CP to Full Power. Valve positions as defined by POP's, SOP's, and COL's SHUTDOWN POSITION: Dcfincd as RCS 200°F and below, not in rcfucling and not at reduced inventory. Valve positions as defined by POP's and SOP's.

POST ACCIDENT POSITION: Defined as SI with Phase A and B isolation. Note: valve position may differ based on the accident in progress (i.e. phase B may not be required).

POST ACCIDENT USAGE: Defined as Design Basis Accident valve usage based on position during long tenn recirculation, assuming no failures. Note: valve position may differ based on the accident in progress, equipment failure, and recommendations during the recovery phase.

NOTES' 1.Penetration elass is described in subsection 5.2.2. 16. Test Fluid "G" signifying Gas indicates either air or nitrogen as test medium.

2.Also tripped closed by high radiation in contaimnent. 17. Seal air via WCCPP, continuously pressurized.

3.Penetration sealed at both ends. 18. May be opened Post Accident if normal path from recirc. pumps not available.

4. Sealed by Isolation Valve Seal Water System. 19. Valves may be closed Post Accident if not in service.

5. "Sealed" by Residual Heat Removal System or recirculation sump fluid. Not a 2U. May be opened Post Accident when the contaimnent pressure is below 5 psig.

"seal system" as defined in 10 CFR 50, Appendix J.

6. "Sealed" by Service Water System fluid. Not a "seal system" as defined in 10 21. Valves may be opened for maintenance.

CFR 50, Appendix J. LLR testing is not required for Appendix J compliance but is required to limit in-leakage to the contaimnent given a postulated breech of SWS integrity during the long-term recovery phase.

7. Sealed by Weld Channel and Contaimnent Penetration 22. Valves outside contaimnent in these lines will automatically isolate for steamline break Pressurization System. or Hi-Hi contaimnent pressure.

8.Non-Automatic Contaimnent Isolation Valves open continuously or intermittently 23. DDV modified due to press. locking to function as a std. gate valve. 885A upstream disc for plant operation (under administrative control). drilled with 3/16"dia. hole, 885B bonnet connection bypasses downstream disc.

9.Valves may be operated as required to support plant operation. 24. Valves may be opened intermittently during Post Accident venting.

10.Thesc scries valvcs have non-redundant phasc A automatic signals and therefore 25. Penetrations identified as H (hot) indicatcs designed with expansion bcllows or are treated as non-automatic contaimnent isolation valves. ex'])ansion coil, C (cold) indicates designed without an expansion bellows or expansion coil.

ll.Isolated when Reactor Coolant Pumps are stopped. 26. Spring-loaded check valves (pressure relieving).

12.Valves opened intermittently to take samples. 27. Flange is double gasketed type, located in refueling canal.

13.Valve opened periodically by the Gas Analyzer. 28. Necessary to LC & de-energize if AC-730 & 731 are de-energized open.

14.0pened intermittently for pressure relief. 29. Control Rm. Annunciator "Personnel hatches not shut" alarm indication provided.

15. Testable only at Cold Shutdown.

"'U m NOTES'

() 30. A Seismic Class I QA CAT M temporary fiber optic penetration flange (TFP)

"'U o m o may be installed in cold shutdown / refueling conditions to satisfy contaimnent oo o isolation function for refueling operations.

eN o 01 31. Once opened to facilitate high head or hot leg recirculation, valves would o0) .j:::o.

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IP 3 FSAR UPDATE TABLE 5.2-3 (Sheet 9 of 9)

COl'lTAINMENT PIPING PENETRATIONS AND VALVING

~ ~ .... ~ . . . IU' ...dL ~ "" * * 'U'YY I ....... I .... I .... ~.u.,,"' ."JL"-JL '"' ..... 'U'U' . . . . . U' .. , , ' "

remain open unless closed to isolate a postulated passive failure during the long-term recovery phase. LLR testing is not required.

32. LLR testing performed to verify adequacy of on-site nitrogen inventory. LLR is not required for Appendix J compliance.

33. Valves remain open to facilitate high head or hot leg recirculation unless closed to isolate a postulated passive failure during the long-term recovery phase. LLR testing is not required.

34. LLRT is not required. Valve / penetration is open during Type A ILR test.

62 of 188

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IP3 FSAR UPDATE TABLE 5.2-4 CONTAINMENT ISOLATION VALVE POSITION INDICATION Control Board Panel Control Board Panel Valve Number Location Red & Green Location (Two-is-True)

Indicating Lights Monitor Lights AC-MOV-743 SNF SB1F AC-MOV-744 SGF&SB1F SB1F CH-AOV-201 SFF & SNF SNF CH-AOV-202 SFF & SNF SNF CH-MOV-222 SFF & SNF SNF RC-AOV-519 SAF SNF RC-AOV-548 SNF SNF RC-AOV-549 SNF SNF RC-AOV-552 SAF SNF AC-FCV-625 SGF & SNF SNF AC-MOV-769 SGF & SNF SNF AC-MOV-784 SGF & SNF SNF AC-MOV-786 SGF & SNF SNF AC-MOV-789 SGF & SNF SNF AC-AOV-791 SGF & SNF SNF AC-AOV-793 SGF & SNF SNF AC-AOV-796 SGF SNF AC-MOV-797 SGF & SNF SNF AC-AOV-798 SGF & SNF SNF SP-AOV-956A Sampling System Panel

• SNF SP-AOV-956B Sampling System Panel

• SNF SP-AOV-956C Sampling System Panel

• SNF SP-AOV-956D Sampling System Panel

• SNF SP-AOV-956E Sampling System Panel

• SNF SP-AOV-956F Sampling System Panel

• SNF SP-AOV-956G Sampling System Panel

• SNF SP-AOV-956H Sampling System Panel

• SNF SP-AOV-959 Sampling System Panel

• SNF VS-FCV-1170 SLF and Fan Room Ctr. Cab.* SNF VS-FCV-1171 SLF and Fan Room Ctr. Cab

• SNF VS-FCV-1172 SLF and Fan Room Ctr. Cab

• SNF VS-FCV-1173 SLF and Fan Room Ctr. Cab

• SNF VS-PCV-1190 SLF and Fan Room Ctr. Cab

• SNF VS-PCV-1191 SLF and Fan Room Ctr. Cab

• SNF VS-PCV-1192 SLF and Fan Room Ctr. Cab

• SNF BD-PCV-1214 SCF SNF DW-AOV-1 SKF SNF DW-AOV-2 SKF SB1F RC-AOV-550 SKF SB1F

• Not located in control room 63 of 188 IPEC00035484 IPEC00035484

IP3 FSAR UPDATE TABLE 5.2-4 (Cont.)

CONTAINMENT ISOLATION VALVE POSITION INDICATION Control Board Panel Control Board Panel Valve Number Location Red & Green Location (Two-is-True)

Indicating Lights Monitor Lights SP-AOV-958 CR Isolation Valve Panel JK1 -

WD-AOV-1610 SKF SNF SI-MOV-850A SB2F&WD Extension * -

SI-MOV-850C SB2F&WD Extension * -

BD-PCV-1214A SCF SNF BD-PCV-1215 SCF SNF BD-PCV-1215A SCF SNF BD-PCV-1216 SCF SNF BD-PCV-1216A SCF SNF BD-PCV-1217 SCF SNF BD-PCV-1217A SCF SNF BD-PCV-1223 Sampling System Panel* SNF BD-PCV-1223A Sampling System Panel

• SNF BD-PCV-1224 Sampling System Panel

• SNF BD-PCV-1224A Sampling System Panel

• SNF BD-PCV-1225 Sampling System Panel

• SNF BD-PCV-1225A Sampling System Panel

• SNF BD-PCV-1226 Sampling System Panel

• SNF BD-PCV-1226A Sampling System Panel

• SNF IA-PCV-1228 SNF SNF CA-PCV-1229 SNF SNF CA-PCV-1230 SNF SNF VS-PCV-1234 SNF SNF VS-PCV-1235 SNF SNF VS-PCV-1236 SNF SNF VS-PCV-1237 SNF SNF WD-AOV-1702 Waste Disposal System Panel* SNF WD-AOV-1705 Waste Disposal System Panel

• SNF WD-AOV-1723 Waste Disposal System Panel

• SNF WD-AOV-1728 Waste Disposal System Panel

• SNF WD-AOV-1786 Waste Disposal System Panel

• SNF WD-AOV-1787 Waste Disposal System Panel

• SNF WD-AOV-1788 SNF SNF WD-AOV-1789 SNF SNF AC-MOV-822A SGF & SB1F SB1F AC-MOV-822B SGF & SB1F SB1F SI-MOV-851A SB2F SB2F NNE-AOV-863 SMF SNF

• Not located In control room 64 of 188 IPEC00035485 IPEC00035485

IP3 FSAR UPDATE TABLE 5.2-4 (Cont.)

CONTAINMENT ISOLATION VALVE POSITION INDICATION Control Board Panel Control Board Panel Valve Number Location Red & Green Location (Two-is-True)

Indicating Lights Monitor Lights SI-MOV-885A SB1F SB1F SI-MOV-885B SB1F SB1F SI-MOV-888A SB1F SB1F SI-MOV-888B SB1F SB1F SI-MOV-1835A SB2F SB2F SI-MOV-1835B SB2F SB2F AC-MOV-1870 SNF SB1F CH-MOV-205 WD Extension * -

CH-MOV-226 WD Extension * -

CH-MOV-250A WD Extension * -

CH-MOV-250B WD Extension * -

CH-MOV-250C WD Extension * -

CH-MOV-250D WD Extension * -

CH-MOV-441 WD Extension * -

CH-MOV-442 WD Extension * -

CH-MOV-443 WD Extension * -

CH-MOV-444 WD Extension * -

SP-MOV-990A WD Extension * -

SP-MOV-990B WD Extension * -

SP-SOV-506 CR Isolation Valve Panel JK1 -

SP-SOV-507 CR Isolation Valve Panel JK1 -

SP-SOV-508 CR Isolation Valve Panel JK1 -

SP-SOV-509 CR Isolation Valve Panel JK1 -

SP-SOV-510 CR Isolation Valve Panel JK1 -

SP-SOV-511 CR Isolation Valve Panel JK1 -

SP-SOV-512 CR Isolation Valve Panel JK1 -

SP-SOV-513 CR Isolation Valve Panel JK1 -

SP-SOV-514 CR Isolation Valve Panel JK1 -

SP-SOV-515 CR Isolation Valve Panel JK1 -

SP-SOV-516 CR Isolation Valve Panel JK1 -

• Not located in control room 65 of 188 IPEC00035486 IPEC00035486

IP3 FSAR UPDATE 5.3 CONTAINMENT VENTILATION SYSTEM 5.3.1 Design Basis 5.3.1.1 Performance Objectives The Containment Ventilation System was designed to accomplish the following:

a) Remove the normal heat loss from all equipment and piping in the Reactor Containment during plant operation and maintain a normal ambient temperature of 130° F or less b) Provide sufficient air circulation and filtering throughout all containment areas to permit safe and continuous access to the reactor containment within two hours after reactor shutdown, assuming defects exist in 1% of the fuel rods.

c) Provide for positive circulation of air across the refueling water surface to assure personnel access and safety during shutdown d) Provide a minimum containment ambient temperature of 50°F during reactor shutdown e) Provide for purging of the containment vessel to the plant vent for dispersion to the environment. The rate of release is controlled by IP3 RECS / ODCM, such that automatic termination of release occurs prior to impacting 10 CFR 20 limits.

f) Provide for depressurization of the containment vessel following an accident. The post-accident design and operating criteria are detailed in Chapter 6 g) Provide ventilation to remove radiogas when steam generator primary man rays are removed h) Provide means for measurement of flow in main plant ventilation exhaust duct In order to accomplish these objectives the following systems were provided:

a) Containment Air Recirculation Cooling and Filtration System b) Control Rod Drive Mechanism Cooling System c) Reactor Compartment Cooling System d) Containment Purge System e) Containment Auxiliary Charcoal Filter System f) Containment Post-Accident Charcoal Filter System (Described in Section 6.4) g) Steam Heating System h) Steam Generator Maintenance Exhaust System.

66 of 188 IPEC00035487 IPEC00035487

IP3 FSAR UPDATE 5.3.1.2 Design Characteristics - Sizing The design characteristics of the equipment required in the Containment for cooling, filtration and heating to handle the normal thermal and air cleaning loads during normal plant operation are presented in Table 5.3-1. In certain cases where engineered safeguards functions also are served by the equipment, component sizing was determined from the heavier duty specifications associated with the Design Basis Accident (DBA), detailed further in Chapter 6.

The fan motors match the power requirements of the fans, which require a maximum power input of 219 horsepower under accident operation. The fan cooler heat removal rate, as a function of the containment pressure, is presented in Section 14.3.6 covering the Containment Integrity Evaluation. For example, this rate at 271°F and 47 psi containment temperature and pressure is 49.0 x 106 Btu/hr per air handling unit. As noted in the Containment Integrity Evaluation, the ability of the Containment Air Recirculation Cooling and Filtration System to function properly in the accident environment was demonstrated by the computer code "HECO."

The code determines the plate fin coil heat removal rate when operating in a saturated steam-air mixture.

5.3.2 System Design 5.3.2.1 Piping and Instrumentation Diagram The containment ventilation, purging and recirculation cooling and filtration systems flow diagram is shown in Plant Drawing 9321-F-40223 [Formerly Figure 6.4-2); The containment ventilation systems and main plant vent were designed as seismic Class I structures.

5.3.2.2 Containment Recirculation Ventilation Air recirculation cooling and filtering during normal operation is accomplished using all five air handling units discharged to a common headered ductwork distribution system to assure adequate flow of filtered and cooled air throughout the Containment. The cooling coils in each air handling unit transfer up to 2.3 x 106 Btu/hr to the Service Water System during normal plant operation and 49.0 x10 6 Btu/hr/FCU in the event of an accident when supplied with 1400 gpm cooling water at 95°F inlet temperature.

Each air handling unit consists of the following equipment arranged so that during normal operation air flows through the unit in the following sequence: cooling coils, centrifugal fan with direct-drive motor, and distribution header.

The fans and motors of these units are equipped with vibration sensors to detect abnormal operating conditions in the early stages of the disturbance. In the event of an accident, the flow path will be diverted automatically by air operated dampers through a compartment containing moisture separators, HEPA filters and charcoal filters. It will then flow through the cooling coils and centrifugal fan and into the distribution header. The normal air flow rate per air handling unit is approximately 70,000 cfm and the post-accident flow rate will be approximately 34,000 cfm, with a 8,000 cfm through the filtration section. Section 6.4.2 provides additional information on the operation of this system.

67 of 188 IPEC00035488 IPEC00035488

IP3 FSAR UPDATE The recirculating ductwork located in the annulus of the Containment Building was provided with spring loaded relief dampers designed to open inward when the external pressure on the ductwork reaches 2 psig. This is discussed in Section 6.4 The Control Rod Drive Cooling System supplements the main containment recirculation system.

The Control Rod Drive Cooling System consists of fans and ductwork to circulate air through the control drive mechanism shroud and discharge it to the main containment volume. Four 1/3 capacity direct driven axial flow fans are used.

5.3.2.3 Containment Purge System The Containment Purge System includes provisions for both supply and exhaust air. The purge system is maintained isolated whenever the plant is above the cold shutdown condition. The supply system includes roughing filters, heating coils, fan, supply penetration with two butterfly valves for bubble tight shutoff, and a purge supply distribution header inside containment. The exhaust system includes exhaust penetration with two butterfly valves identical to those above, exhaust ductwork, filter bank with roughing, HEPA and charcoal filters, fans and exhaust vent.

Provision was made to measure isokinetic flows at the radiation monitor using pitot tubes. The purge system flow rate is 28,000 cfm; however, the isolation valves will be shut prior to going above cold shutdown and will remain closed during normal operation. The quick closing purge isolation valves are capable of closing within two seconds of receipt of the accident signal. The weld channel and penetration pressurization system pressurizes the space between the purge valves and therefore serves as a continuous on-line monitoring system for valve leakage.

During power operation, containment integrity is maintained with no release from the containment ventilation system to the atmosphere. Prior to purging the Containment, air particulate and gas monitor indications of the closed containment activity levels are used as a guide to making routine releases from the Containment. During power operation, the containment air particulate and gas monitor indications help determine the desirability of using either one or both of two auxiliary particulate and charcoal filter units installed in the Containment primarily for pre-access cleanup.

When the containment purging for access following reactor shutdown is in progress, releases from the plant vent are continuously monitored for radiogas and particulates and sampled for iodine and tritium. A wide range plant vent gas monitor (Section 11.2.3.1) provides continuous indication of noble gas releases passing through the plant vent to the atmosphere.

5.3.2.4 Isolation Valves The purge supply and exhaust ducts butterfly valves, both inside and outside the containment, are closed during power operation. The spaces between the closed valves are pressurized with air by the Penetration and Weld Channel Pressurization System. The valves were designed for rapid automatic closing by the containment isolation signal (derived from any automatic safety injection signal), or upon a signal of high activity level within the Containment in the event of a radioactivity release when the purge line is open.

5.3.2.5 Containment Pressure Relief Line The normal pressure changes in the Containment during reactor power operation will be handled by the containment pressure relief line. This line is equipped with three quick-closing butterfly type isolation valves, one inside and two outside the Containment. The valves will be 68 of 188 IPEC00035489 IPEC00035489

IP3 FSAR UPDATE automatically actuated to the closed position by the containment isolation signal, or by a containment high radioactivity signal. The two intra-valve spaces are pressurized with air by the Penetration and Weld Channel Pressurization System when the valves are closed. The pressure relief line discharges through roughing, HEPA, and charcoal filters to the plant vent.

While the valves are fully capable of closing from a 60° open position during accident conditions, mechanical stops prevent the valves from opening more than 40° (90° = full open).

5.3.2.6 Steam Generator Maintenance Exhaust System Steam generator maintenance ventilation is accomplished by use of two 3000 cfm fans driven by 5 hp motors. These fans connect to 14" diameter exhaust ducts, which allow maintenance on the steam generators when the manways are removed. The fans exhaust into the containment purge exhaust duct.

5.3.2.7 Pressurizer Relief Tank Venting During shutdown conditions, the potential exists for radioactive gases to be vented from the Pressurizer Relief Tank. These gases are therefore routed to the containment purge exhaust duct where their radioactive content can be monitored (see Section 11.2).

The system uses a jet eductor, using station air to vent the tank. The system is shut down during normal operation.

69 of 188 IPEC00035490 IPEC00035490

IP3 FSAR UPDATE TABLE 5.3-1 PRINCIPAL COMPONENT DATA

SUMMARY

Units Required Units for System Installed Unit Capacity Normal Operation Containment Recirculation Demister 5 8,000 cfm 0 Cooling Coils - Normal 5 2.3 x 106 Btulhr 5 Cooling Coils - DBA 5 49.0 x 106 Btulhr 0 HEPA Filters 5 8,000 cfm 0 Fans 5 70,000* cfm 5 Fans Pressure - Normal 6.3 in H2O Fan Motors (440 V, 3 phase) 5 225 hp 5 DBA Charcoal Filters 5** 8,000 cfm 0 Temperature Switches 30**

Control Rod Drive Mechanism Cooling Fans, Standard Conditions 4 15,000 cfm 3 Fan Pressure 5-1/2 in H2O Fan Motors 4 25 hp 3 Reactor Compartment Cooling Part of CB Recirculation 12,000 cfm System Refueling Canal Air Sweep Part of CB Recirculation 17,5000 cfm System 70 of 188 IPEC00035491 IPEC00035491

IP3 FSAR UPDATE TABLE 5.3-1 (Cant.)

PRINCIPAL COMPONENT DATA

SUMMARY

Units Required Units for System Installed Unit Capacity Normal Operation Containment Ventilation/Purge Supply Fans, Standard Conditions 1 40,000 cfm Optional Approximately Fan Pressure 2.75 in H2 O Fan Motors 1 40 hp Pre-heat Coils 1 Set Optional Air Filters, Roughing 1 40,000 cfm 1 Exhaust Fans,* Standard Conditions 2 70,000 cfm** Optional Fan Pressure Fan Motors 2 150 hp Plenums 2 40,000 cfm HEPA Filters 1 Bank 40,000 cfm Optional Roughing Filters 1 Bank 40,000 cfm Optional Charcoal Filters 1 Bank 40,000 cfm Optional Containment Auxiliary Charcoal Filters Fans, Standard Conditions 2 8,000 cfm Optional Fan Pressure 4.75 in H2 0 Fan Motors 2 10 hp Filters; Roughing, HEPA 2 8,000 cfm Optional and Charcoal Filters

• Note: The two exhaust fans are used interchangeably or as backup for:

1. Ventilation of Primary Auxiliary Building (70,000 cfm)

2. Containment Building Purge System (40,000 cfm)

• • Note: Normal System Flow for Containment Building Purge Exhaust is 28,000 cfm.

71 of 188 IPEC00035492 IPEC00035492

IP3 FSAR UPDATE TABLE 5.3-1 (Cant.)

PRINCIPAL COMPONENT DATA

SUMMARY

Units Required Units for System Installed Unit Capacity Normal Operation Steam Heating Heaters, 25 psig steam 2 400,000 Btu/hr each Optional Steam Generator Maintenance Exhaust System Centrifugal Fan 2 3000 cfm 2 Fan Pressure Fan Motors 2 5 hp 2 Containment Building Pressure Relief Fan, Standard Conditions 1 1500 cfm Optional Fan Pressure Fan Motor 5 hp Filters; Roughing, HEPA 1 1500 cfm Optional and Charcoal Filters Note: The operating configuration for the Containment Building Pressure Relief system involves limiting the three containment isolation valves to a minimum position of 40° open. This causes a decrease in system flow.

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IP3 FSAR UPDATE 73 of 188 IPEC00035494 IPEC00035494

IP3 FSAR UPDATE 74 of 188 IPEC00035495 IPEC00035495

IP3 FSAR UPDATE 75 of 188 IPEC00035496 IPEC00035496

IP3 FSAR UPDATE 76 of 188 IPEC00035497 IPEC00035497

IP3 FSAR UPDATE 77 of 188 IPEC00035498 IPEC00035498

IP3 FSAR UPDATE 78 of 188 IPEC00035499 IPEC00035499

IP3 FSAR UPDATE 79 of 188 IPEC00035500 IPEC00035500

IP3 FSAR UPDATE 80 of 188 IPEC00035501 IPEC00035501

IP3 FSAR UPDATE 5.5 CONTAINMENT PARAMETERS The description of the instrumentation system included in the Indian Point 3 design for remote monitoring of post-accident conditions within the primary containment is presented in Appendix 6F. Non-nuclear process instrumentation of the containment is described in Section 7.5.

Containment Building Pressure The containment pressure is transmitted to the main control board for post accident monitoring.

Six transmitters, two in each of three safety channels, are installed outside the containment to prevent potential missile damage. The pressure is indicated on the main control board; the range is -5 psig to 75 psig.

In addition, monitoring of the containment building pressure during and following an accident is effected by two Safety Category I redundant systems. Pressure signals are obtained at the pipe penetration area and brought to transmitters outside containment. These same signals are transmitted to the two-recorders at the control room recorder cabinet. Continuous monitoring of containment pressure is possible in the -5 to 200 psig range. Power requirements for the two systems are met from vital instrument buses. The installation of cable and conduit is consistent with separation criteria, as outlined in Section 8.4.

Containment Building Water Level Monitoring There are three sumps in the Containment Building: Reactor Pit Sump, Recirculation Sump and Containment Sump. Associated with the Recirculation Sump and Containment Sump, there are two redundant, separately channeled and powered level measurement loops. Associated with the Reactor Pit is a level sensor, alarmed on the Control Room Supervisory Panel. These provide continuous level and alarm indication in the Control Room. Additionally, a water level transmitter installed at the top of Containment Sump will provide a Containment Sump overflow alarm indication in the Control Room.

Containment Building Hydrogen Concentration Hydrogen concentration indication is provided by a measuring system which consists of the following: redundant analyzers and continuously recording two (2) single pen recorders. The recorders are located in the control room and the analyzers are located in the pipe penetration area of the fan house. Samples are drawn from containment recirculation fans via the retired post-accident sample system and returned to the general area of the containment building.

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IP3 FSAR UPDATE CHAPTER 6 ENGINEERED SAFETY FEATURES 6.1 GENERAL DESIGN CRITERIA Criteria applying in common to all engineered safety features are given in Section 6.1.1. Criteria which are related to engineered safety features, but which are applicable to specific features or systems, are listed and cross referenced in Section 6.1.2.

The engineered safety features are discussed in detail in this Chapter. In each section a separate safety feature is described and evaluated. In the evaluation section for each engineered safety feature, a single failure evaluation is provided which delineates the components of that safety feature system and the interconnected auxiliary systems that must function for the proper operation of that engineered safety feature. An examination of these tables shows that some components of the Residual Heat Removal System, Component Cooling Water System, and the Service Water Systems are necessary for proper operation of the Engineered Safety Features. These systems and their components are discussed in Sections 9.3 and 9.6; the instrumentation associated with these systems is also discussed in the referenced sections. Since the auxiliary system components, both inside and outside the containment, and their instrumentation and power systems are not required for actuation of the engineered safety features, neither IEEE-279 nor the General Design Criteria apply.

The General Design Criteria presented and discussed in this section are those which were in effect at the time when Indian Point 3 was designed and constructed. These general design criteria, which formed the bases for the Indian Point 3 design, were published by the Atomic Energy Commission in the Federal Register of July 11, 1967, and subsequently made a part of 10 CFR 50.

The Authority has completed a study of compliance with 10 CFR Parts 20 and 50 in accordance with some of the provisions of the Commission's Confirmatory Order of February 11, 1980. The detailed results of the evaluation of compliance of Indian Point 3 with the General Design Criteria established by the Nuclear Regulatory Commission (NRC) in 10 CFR 50 Appendix A, and in effect at the time of study, were submitted to NRC on August 11, 1980, and approved by the Commission on January 19,1982. These results are presented in Section 1.3.

6.1.1 Engineered Safety Features Criteria Engineered Safety Features Basis for Design Criterion: Engineered safety features shall be provided in the facility to back up the safety provided by the core design, the reactor coolant pressure boundary, and their protection systems. Such engineered safety features shall be designed to cope with any size reactor coolant piping break up to and including the equivalent of a circumferential rupture of any pipe in that boundary, assuming unobstructed discharge from both ends. (CDC 37 of 7/11/67)

The design, fabrication, testing and inspection of the core, reactor coolant pressure boundary and their protection systems give assurance of safe and reliable operation under all anticipated normal, transient, and accident conditions. However, engineered safety features are provided in the facility to back up the safety provided by these components. These engineered safety 1 of 215 IPEC00035610 IPEC00035610

IP3 FSAR UPDATE features were designed to cope with any size reactor coolant pipe break, up to and including the circumferential rupture of any pipe, assuming unobstructed discharge form both ends, and to cope with any steam or feedwater line break, up to and including the main steam or feedwater headers.

Limiting the release of fission products from the reactor fuel is accomplished by the Safety Injection System which, by cooling the core, keeps the fuel in place and substantially intact, and limits the metal water reaction to an insignificant amount.

The Safety Injection System consists of high and low head centrifugal pumps driven by electric motors, and passive accumulator tanks which are self actuated and which act independently of any actuation signal or power source.

The release of fission products from the containment is limited in three ways:

Blocking the potential leakage paths from the containment. This is accomplished by:

A steel-lined, reinforced concrete Reactor Containment with testable, doubly sealed penetrations and most liner weld channels, the spaces of which are continuously pressurized above accident pressure, and which form a virtually leak-tight barrier to the escape of fission products should a loss-of-coolant accident occur.

Isolation of process lines by the Containment Isolation System which imposes double barriers in each line which penetrates the containment except for lines utilized during the accident. An Isolation Valve Seal Water System provides a water or nitrogen seal at the isolation valves thus sealing some of the pipes penetrating the containment.

Reducing the fission product concentration in the containment atmosphere. This is accomplished by:

a) Containment Air recirculation filters which provide for rapid removal of particles and iodine vapor from the containment atmosphere.

b) Chemically treated spray which removes elemental iodine vapor from the containment atmosphere by washing action.

Reducing the containment pressure and thereby limiting the driving potential for fission product leakage. This is accomplished by cooling the containment atmosphere by the following independent systems:

a) Containment Spray System b) Containment Air Recirculation and Cooling System Reliability and Testability of Engineered Safety Features Criterion: All engineered safety features shall be designed to provide such functional reliability and ready testability as is necessary to avoid undue risk to the health and safety of the public. (GDC 38 of 7/11/67) 2 of 215 IPEC00035611 IPEC00035611

IP3 FSAR UPDATE A comprehensive program of plant testing was formulated for all equipment, systems and system controls vital to the functioning of engineered safety features. The program consists of performance tests of individual pieces of equipment in the manufacturer's shop, integrated tests of the system as a whole, and periodic tests of the actuation circuitry and mechanical components to assure reliable performance, upon demand, throughout the plant lifetime.

The initial tests of individual components and the integrated test of the system as a whole complemented each other to assure performance of the system as designed and to demonstrate the proper operation of the actuation circuitry.

Routine periodic testing of the engineered safety features components is performed as specified in the Technical Specifications.

Missile Protection Criterion: Protection for engineered safety features shall be provided against dynamic effects and missiles that might result from plant equipment failures. (GDC 40 of 7/11/67)

A Loss-of-Coolant Accident or other plant equipment failures might result in dynamic effects or missiles. For engineered safety features which are required to ensure safety in the event of such an accident or equipment failure, protection is provided primarily by the provisions which are taken in the design to prevent the generation of missiles. In addition, protection is also provided by the layout of plant equipment or by missile barriers in certain cases. See Chapter 5 for a discussion of missile protection. The dynamic effects associated with postulated pipe breaks in the Primary Coolant System (hot legs, cold legs, crossover legs) need not be a design basis (NCR SER dated March 10, 1986).

Injection paths leading to unbroken reactor coolant loops are protected against damage as a result of the maximum reactor coolant pipe rupture by layout and structural design considerations. Injection lines penetrate the main missile barrier, which is the crane wall, and the injection headers are located in the missile-protected area between the crane wall and the containment wall. Individual injection lines, connected to the injection header, pass through the barrier and then connect to the loops. Separation of the individual injection lines is provided to the maximum extent practicable. Movement of the injection lines, associated with rupture of a reactor coolant loop, is accommodated by line flexibility and by the design of the pipe supports such that no damage outside the missile barrier is possible.

The containment structure is capable of withstanding the effects of missiles originating outside the containment, and which might be directed toward it, so that no Loss-of-Coolant Accident can result from these missiles.

All hangers, stops and anchors were designed in accordance with ANSI B31.1 Code for Pressure Piping and ACI 318 Building Code Requirements for Reinforced Concrete which provide minimum requirements on material, design and fabrication with ample safety margins for both dead and dynamic loads over the life of the equipment. Additional information on the design and re-analyses of hangers, stops and anchors is presented in Section 16.3.

Where necessary to prevent pipe whip, restraints were installed with the proper arrangement and spacing to prevent a plastic hinge mechanism from forming as a result of the forces associated with a pipe rupture. Restraint spacing was determined by calculation of the unsupported pipe length resulting in a plastic hinge formation for two basic support arrangement 3 of 215 IPEC00035612 IPEC00035612

IP3 FSAR UPDATE and break location cases. Both slot and guillotine breaks were considered. Slot breaks are defined as instantaneous openings in the pipe parallel to the axis of the pipe with an opening length twice the length of the nominal pipe diameter and with an opening area equal to the area of the pipe interior cross-section.

Slot breaks were assumed to occur anywhere in the piping system, including fittings. Guillotine breaks are defined as instantaneous severance of the pipe cross-section and are assumed to occur any point of discontinuity in the piping system (such as valves, fittings and elbows). The pipe break loads were determined from P' = PoA where Po = system pressure A = inside cross sectional area, except for the main steam lines downstream of the flow limiting device, where force resultants are limited by the restriction of the flow limiting device. Such loads for the steam lines were taken as P' = 340 kips; and for the feedwater lines, P' = 200 kips. For both slot and guillotine breaks, restraints were spaced such that plastic hinge mechanisms cannot form in the piping system which would permit unrestrained rotation of the piping.

The restraints were designed such that the maximum applied load or stress be less than the lesser of the yield strength of the material or 0.67 times the rated ultimate load capacity of the support. High strength cable restraints were designed such that the maximum applied load be less than 0.4 times the rated ultimate load capacity of the cable. In those instances where the integrity of the restraint is also dependent on reinforced concrete anchorage, the concrete behavior limits are in accordance with ACI-318-63, Part IV-B, requirements and bearing stress is limited to 0.8 f'e.

Vital equipment is protected from pipe whip by locating restraints on nearby high pressure lines such that the two free ends of a broken pipe cannot reach the equipment.

The plant arrangement provides the basic protection against pipe whip. The four loops of the primary coolant system are spaced to the maximum extent possible; the crane wall protects the reactor compartment from pipe whip in the annulus; pipe lines are run radially outward from the reactor compartment. Wherever pOSSible, redundant engineered safeguards piping is physically separated so that a failure of one pipe and subsequent whipping cannot cause the failure of the second pipe. Where physical separation is impossible, for instance the Accumulator Tanks' discharge piping, both pipes are restrained in such a way that a plastic hinge cannot form in case of a double ended rupture.

Containment fan cooler units are separated from high pressure pipe lines by the floor at Elev.

68' -0".

Small lines are treated no differently from large lines in so far as containment isolation, separation, pipe whip protection, etc. Separation is provided where whipping of larger lines would otherwise result in damage to many small lines.

4 of 215 IPEC00035613 IPEC00035613

IP3 FSAR UPDATE Small lines having significant internal pressure are supported and restrained in a manner that would preclude any failure of the containment vessel from the failure of the small line. In addition, see Section 5.2 for the containment isolation provisions for these lines.

Engineered Safety Features Performance Capability Criterion: Engineered safety features, such as the emergency core cooling system and the containment heat removal system, shall provide sufficient performance capability to accommodate the failure of any single active component without resulting in undue risk to the health and safety of the public. (GDC 41 of 7/11/67)

Each engineered safety feature provides sufficient performance capability to accommodate any single failure of an active component and still function in manner to avoid undue risk to the health and safety of the public.

The extreme upper limit of public exposure is taken as the levels and time periods presently outlined in 10 CFR 100, i.e., 300 rem to the thyroid in two hours at the exclusion radius and 300 rem to the thyroid over the duration of the accident at the low population zone distance. The accident condition considered is the hypothetical case of a release of fission products per TID 14844. Also, the total loss of all outside power is assumed concurrently with this accident. With all engineered safety features systems functioning at full capacity, the offsite exposure would be within 10 CFR 20 limits.

Under the above accident conditions, the Containment Air Recirculation Cooling and Filtration System and the Containment Spray System are designed and sized so that either system operating with partial effectiveness is able to supply the necessary post-accident cooling capacity to assure the maintenance of containment integrity, that is, keeping the pressure below design pressure at all times, assuming that the core residual heat is released to the containment as steam. Partial effectiveness is defined as operation of a system with at least one active component failure. Both systems together, each operating with partial effectiveness, are capable of providing the necessary post-accident iodine removal such that the resulting off-site exposures are within the guidelines of 10 CFR 100.

Engineered Safety Features Components Capability Criterion: Engineered safety features shall be designed so that the capability of these features to perform their required function is not impaired by the effects of a Loss-of-Coolant Accident to the extent of causing undue risk to the health and safety of the public. (GDC 42 of 7/11/67)

Instrumentation, pumps, fans, filters, cooling units, valves, motors, cables and penetrations located inside the containment were selected to meet the most adverse accident conditions to which they may be subjected. These items are either protected from containment accident conditions or were designed to withstand, without failure, exposure to the worst combination of temperature, pressure, and humidity expected during the required operational period.

The Safety Injection System pipes serving each loop are anchored at the crane wall, which constitutes the missile barrier in each loop area, to restrict potential accident damage to the portion of piping beyond this point. The anchorage was designed to withstand, without failure, the thrust force of any branch line, severed from the reactor coolant pipe and discharging fluid to the atmosphere; and to withstand a bending moment equivalent to that which produces failure 5 of 215 IPEC00035614 IPEC00035614

IP3 FSAR UPDATE of the piping under the action of free discharge to atmosphere or motion of the broken reactor coolant pipe to which the injection pipes are connected. This prevents possible failure at any point upstream from the support point including the branch line connection into the piping header.

Accident Aggravation Prevention Criterion: Protection against any action of the engineered safety features which would accentuate significantly the adverse after-effects of a loss of normal cooling shall be provided. (GDC 43 of 7/11/67)

The reactor is to be maintained subcritical following a pipe rupture accident. Introduction of borated cooling water into the core results in a net negative reactivity addition. The control rods are inserted and remain inserted.

The supply of water by the Safety Injection System to cool the core cladding reduces the potential for significant metal-water reaction (less than 1.0%).

The delivery of cold safety injection water to the reactor vessel following accidental expulsion of reactor coolant does not cause further loss of integrity of the Reactor Coolant System boundary.

Sharing of Systems Criterion: Reactor facilities may share systems or components if it can be shown that such sharing will not result in undue risk to the health and safety of the public (GDC 4 of 7/11/67)

The residual heat removal pumps and heat exchangers serve dual functions. Although the normal duty of the residual heat exchangers and residual heat removal pumps is performed during periods of reactor shutdown, during all plant operating periods these residual heat removal pumps are aligned to perform the low head safety injection function. In addition, during the recirculation phase of a Loss-of-Coolant Accident, the residual heat exchangers of this system perform the core cooling function and the containment cooling function as part of the Containment Spray System, and the residual heat removal pumps, which are part of the external recirculation loop, provide back-up capability to the recirculation pumps which comprise part of the internal recirculation loop.

Demonstration checking of the system, performed as dictated by the Technical Specifications, provides assurance of correct system alignment for the safety injection function of the components.

During the injection phase, the safety injection pumps do not depend on any portion of other systems. During the recirculation phase, if Reactor Coolant System pressure stays high due to a small break accident, suction to the safety injection pumps is provided by the internal recirculation pumps, and can also be provided by the Residual Heat Removal pumps.

The Containment Air Recirculation and Filtration System also serves the dual function of containment cooling during normal operation and containment cooling after an accident. Since the method of operation for both cooling functions is the same, the dual aspect of the system does not affect its function as an engineered safety feature.

6 of 215 IPEC00035615 IPEC00035615

IP3 FSAR UPDATE The steam supply and city water systems at the Indian Point site were shared by all three reactor facilities. However, independent steam supply and city water systems have been installed at Indian Point 3 (See Chapter 9); the city water system for Indian Point 2 is presently used by Indian Point 3 as a backup supply. The steam supply and city water systems are used for the following purposes:

a) Steam for unit heaters for standby heating.

b) Steam to valved hose connections for maintenance purposes.

c) Water to emergency showers.

d) Water to hose connections for maintenance purposes.

e) (Deleted) f) Water supply to fire protection tanks.

g) Water supply for make-up demineralizers in Condensate Polishing Facility (CPF).

h) Redundant source of makeup water to the spent fuel pit.

i) Backup water supply to Charging Pumps' Fluid Drive Coolers.

6.1.2 Related Criteria The following are criteria which, although related to all engineered safety features, are more specific to other plant features or systems, and therefore are discussed in other sections, as listed:

Title of Criterion (7/11/67 issue) Reference Quality Standards (GDC 1) Chapter 4 Performance Standards (GDC 2) Chapter 4 Records Requirements (GDC 5) Chapter 4 Instrumentation and Control Systems (GDC 12) Chapter 7 Engineered Safety Features Protection Systems (GDC 15) Chapter 7 Emergency Power (GDC 39 and GDC 24) Chapter 8 6.2 SAFETY INJECTION SYSTEM 6.2.1 Design Basis The General Design Criteria presented and discussed in this section are those which were in effect at the time when Indian Point 3 was designed and constructed. On November 22, 1965, the Atomic Energy Commission (AEC) published and requested comments on Proposed 7 of 215 IPEC00035616 IPEC00035616

IP3 FSAR UPDATE General Design Criteria which were developed to assist in the evaluation of applications for nuclear power plant construction permits. On July 11, 1967, a revised set of General Design Criteria were published for comment. The revision reflected extensive public comments, suggestions from meetings with the Atomic Industrial Forum (AIF) and review within the AEC.

In the July to October 1967 time frame, AI F Incorporated assembled nuclear industry comments and transmitted to the AEC revised wording of the 1967 Draft General Design Criteria along with a description of the changes. It was the AIF version of the 1967 General Design Criteria which formed the bases of the Indian Point 3 design and are discussed in this section. The AEC subsequently revised the 1967 version of the General Design Criteria and incorporated them into 10 CFR 50, Appendix A in 1971.

The Authority has completed a study of compliance with 10 CFR Parts 20 and 50 in accordance with some of the provisions of the Commission's Confirmatory Order of February 11, 1980. The detailed results of the evaluation of compliance of Indian Point 3 with the General Design Criteria presently established by the Nuclear Regulatory Commission (NRC) in 10 CFR 50 Appendix A, were submitted to NRC on August 11, 1980, and approved by the Commission on January 19, 1982. These results are presented in Section 1.3.

Emergency Core Cooling System Capability Criterion 44: An Emergency Core Cooling System with the capability for accomplishing adequate emergency core cooling shall be provided. This core cooling system and the core shall be designed to prevent fuel and clad damage that would interfere with the emergency core cooling function and to limit the clad metal water reaction to acceptable amounts for all sizes of breaks in the reactor coolant piping up to the equivalent of a double-ended rupture of the largest pipe. The performance of such emergency core cooling system shall be evaluated conservatively in each area of uncertainty.

Adequate emergency core cooling is provided by the Safety Injection System (which constitutes the Emergency Core Cooling System) whose components operate in three modes. These modes are delineated as passive accumulator injection, active safety injection and residual heat removal recirculation.

The system assures that the core will remain intact and in place with its essential heat transfer geometry preserved following a rupture in the Reactor Coolant System. It also assures that the extent of metal-water reaction is limited such that the amount of hydrogen generated from this source in combination with that from other sources, is tolerable in the Containment.

This capability is provided during the simultaneous occurrence of a Design Basis Earthquake.

This protection is afforded for:

1) All pipe break sizes up to and including the hypothetical instantaneous circumferential rupture of a reactor coolant loop, assuming unobstructed discharge from both ends,

2) or loss of coolant associated with the rod ejection accident,

3) or steam generator tube rupture.

8 of 215 IPEC00035617 IPEC00035617

IP3 FSAR UPDATE The primary function of the emergency Core Cooling System (ECCS) for the ruptures described is to remove the stored and fission product decay heat from the core such that fuel damage to the extent that would impair effective cooling of the core is prevented. This implies that the core remain intact and in place with its essential heat transfer geometry preserved. To assure effective cooling of the core, limits on peak clad temperature and local metal-water reaction will not be exceeded. It has been demonstrated in the Westinghouse Rod Burst Program that for conditions within the area of safe operation, fuel rod integrity is maintained.

To limit the production of hydrogen in the Containment, the overall metal water reaction is limited to 1%.

In evaluating ECCS performance, consideration was given to core geometry distortion caused by swelling or fuel rod bursting.

For any rupture of a steam pipe and the associated uncontrolled heat removal from the core, the Safety Injection System (SIS) adds shutdown reactivity so that with a stuck rod, no offsite power and minimum engineered safety features, there is no consequential damage to the Reactor Coolant System and the core remains in place and intact.

Redundancy and segregation of instrumentation and components are incorporated to assure that postulated malfunctions will not impair the ability of the system to meet the design objectives. The system is effective in the event of loss of normal station auxiliary power coincident with the loss of coolant, and is tolerant of failures of any single component or instrument channel to respond actively in the system. During the recirculation phase, the system is tolerant of a loss of any part of the flow path since backup alternative flow path capability is provided.

The ability of the Safety Injection System to meet its capability objectives is presented in Section 6.2.3. The analysis of the accidents is presented in Chapter 14.

Inspection of Emergency Core Cooling System Criterion 45: Design provisions shall, where practical, be made to facilitate inspection of all physical parts of the Emergency Core Cooling System, including reactor vessel internals and water injection nozzles.

Design provisions are made to the extent practical in order to facilitate access to the critical parts of the reactor vessel internals, pipes, valves and pumps for visual or boroscopic inspection for erosion, corrosion and vibration wear evidence and for non-destructive test inspection where such techniques are desirable and appropriate as detailed in Section 6.2.5.

Testing of Emergency Core Cooling System Components Criterion 46: Design provisions shall be made so that components of the Emergency Core Cooling System can be tested periodically for operability and functional performance.

The design provides for periodic testing or active components of the Safety Injection System for operability and functional performance as detailed in Section 6.2.5.

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IP3 FSAR UPDATE Power sources are arranged to permit individual actuation of each active component of the Safety Injection System.

The safety injection pumps can be tested periodically during plant operation using the minimum flow recirculation lines provided. The residual heat removal pumps are used every time the residual heat removal loop is put into operation and can be tested periodically. All remote operated valves can be exercised and actuation circuits can be tested during routine plant maintenance.

Testing of Emergency Core Cooling System Criterion 47: Capability shall be provided to test periodically the operability of the Emergency Core Cooling System up to a location as close to the core as is practical.

An integrated system test is performed when the plant is cooled down and the residual heat removal loop is in operation. This test would not introduce flow into the Reactor Coolant System but would demonstrate the operation of the valves, pump circuit breakers, and automatic circuitry upon initiation of safety injection.

Level and pressure instrumentation are provided for each accumulator tank, and accumulator tank pressure and level are continuously monitored during plant operation. Flow from the tanks can be checked at any time using test lines.

The accumulators and the safety injection piping up to the final isolation valve are maintained full of borated water at boron concentrations consistent with the accident analysis while the plant is in operation. The accumulators and injection lines are refilled with borated water as required by using the safety injection pumps to recirculate refueling water through the injection headers. A small bypass line and a return line are provided for this purpose.

Flow in each of the high head injection branch lines and in the main flow line for the residual heat removal pumps is monitored by a flow indicator.

Pressure instrumentation is also provided for the main flow paths of the high head and residual heat removal pumps.

Testing of Operational Sequence of Emergency Core Cooling System Criterion 48: Capability shall be provided to test initially, under conditions as close as practical to design, the full operational sequence that would bring the Emergency Core Cooling System into action, including the transfer to alternate power sources.

The design provides for capability to test, to the extent practical, the full operational sequence up to the design conditions for the Safety Injection System to demonstrate the state of readiness and capability of the system. Details of the operational sequence testing are presented in Section 6.2.5, Tests and Inspections.

Engineered Safety Features The Engineered Safety Features are discussed in detail herein. In each section of this Chapter 6 a separate safety feature is described and evaluated. In the evaluation section for each Engineered Safety Feature, a single failure table is provided which lists the components of that safety feature system and the interconnected auxiliary systems that must function for the proper 10 of 215 IPEC00035619 IPEC00035619

IP3 FSAR UPDATE operation of that Engineered Safety Feature. An examination of these tables shows that some components of the Residual Heat Removal System, Component Cooling System, and Service Water System are necessary for proper operation of the Engineered Safety Features. These systems and their components are discussed in Section 9.3 and 9.6. The instrumentation associated with these systems is also discussed in those sections. As the auxiliary system components outside the Containment, as well as those inside the Containment, their instrumentation and power systems are not required for actuation of the Engineered Safety features; neither IEEE-279 nor the General Design Criteria apply.

Codes and Classifications Table 6.2.1 tabulates the codes and standards to which the Safety Injection System components were designed.

Service Life All portions of the system located within the Containment were designed to operate without benefit of maintenance and without loss of functional performance for the duration of time the component is required.

6.2.2 System Design and Operation

System Description

Adequate emergency core cooling following a Loss-of-Coolant Accident is provided by the Safety Injection System as shown in Plant Drawings 9321-F-27353 and -27503 [Formerly Figures 6.2-1A & 6.2~1 B]. The system components operate in the following possible modes:

1) Injection of borated water by the passive accumulators.

2) Injection of borated water from the Refueling Water Storage Tank with the safety injection pumps. (NOTE: Technical Specification Amendment 139 eliminates the requirement to maintain a boron injection tank.)

3) Injection by the residual heat removal pumps also drawing borated water from the Refueling Water Storage Tank.

4) Recirculation of spilled reactor coolant, injected water and Containment Spray System drainage back to the reactor from the recirculation sump by the recirculation pumps. (The residual heat removal pumps provide backup recirculation capability.)

The initiation signal for core cooling by the safety injection pumps and the residual heat removal pumps is the safety injection signal which is actuated by any of the following:

Low pressurizer pressure (2/3)

High containment pressure (2/3, High Pressure)

High differential pressure between any other two steam generators (2/3) 11 of 215 IPEC00035620 IPEC00035620

IP3 FSAR UPDATE After time delay (maximum of 6 seconds): high steam flow in any two of the four steam lines (1/2 per line) coincident with low Tav9 (2/4) or low steam pressure (2/4)

Manual Actuation High-High containment pressure (two sets of 213, High-High pressure) [energize to actuate]

In the Technical Specifications, limits are set on minimum number of operable channels and required plant status for all reactor protection and ESF instrumentation.

Injection Phase The principal components of the Safety Injection system which provide emergency core cooling immediately following a loss of coolant are the accumulators (one for each loop), the three safety injection (high head) pumps and the two residual heat removal (low head) pumps. The safety injection and residual heat removal pumps are located in the Primary Auxiliary Building.

The accumulators, which are passive components, discharge into the cold legs of the reactor coolant piping when pressure decreases below the N2 cover gas operating pressure (approximately 650 psig), thus rapidly assuring core cooling for large breaks. They are located inside the Containment, but outside the crane wall; therefore, each is protected against possible missiles.

The safety injection signal starts the safety injection and residual heat removal pumps and opens the Safety Injection System isolation valves (certain valves have their motor leads disconnected and are locked open). The valves on Plant Drawings 9321-F-27353 and -27503

[Formerly Figures 6.2-1A & -B] marked with a "s" receive the safety injection signal.

Separate and independent key-lock switches one for each SI train are provided in series to each of the auto SI actuation relays to allow manual blocking of the automatic Engineered Safeguards System actuation when the unit is in cold shutdown.

The operation of the key-lock switches into the "defeat" position will activate the existing separate annunciation for each train (Safeguard Train "A" in test and Safeguard Train "B" in test) and separate status lights (one for each train) in the Control Room. While the operator can deactivate the alarm, the individual status lights and the alarm windows will stay lit as long as the key-lock switches are in the defeat position.

The considerations involved insure that:

1) The operation of the key-lock switch to defeat the auto SI is normally carried out during the plant conditions which do not require the actuation of auto SI.

The key-lock switch will be used only during normal plant operation, with the plant in the cold shutdown condition. The Technical Specifications do not require the operability of the SI system or any of its components during the cold shutdown conditions.

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IP3 FSAR UPDATE

2) The operation of the key-lock switch to defeat auto SI is also permitted following an SI activation if the normal method or resetting SI is unavailable. This action is required to restore control of plant equipment to the operators.

3) Annunciation devices are provided to augment the administrative procedures.

The operation of the key-lock switch will activate the individual annunciations and individual status lights. During the time auto SI is in the "defeat" position, the "alarm windows" and the status lights will stay lit.

The safety injection pumps (high head) deliver borated water to two separate discharge headers. The flow from each header can be injected into each of the three available cold legs (one of four cold leg lines per header has been permanently isolated by locking closed valves SI-856A on 2" Line #56 and SI-856F on 1-%" Line #754, as evaluated in Reference 2) and one hot leg of the Reactor Coolant System. Isolation valves in each of the three available cold leg injection lines are open and valves in the hot leg injection lines are closed during normal plant operation. The hot leg injection lines are provided for later use during hot leg recirculation following a reactor coolant pressure boundary break.

One high head injection header contains the retired-in-place Boron Injection Tank (BIT), which formerly contained concentrated boric acid for rapid insertion of negative reactivity in the safety injection mode. A modification replaced the contents of the BIT with water from the Refueling Water Storage Tank (RWST), (Reference 3).

NOTE: Technical Specification Amendment 139 eliminates the requirement to maintain a BIT.

However, the BIT is an in-line, passive component of the Safety Injection System, therefore it is only "functionally eliminated," but not physically. Furthermore, in the event of a safety injection scenario, the BIT will continue to "function" in a passive mode, to convey refueling water to the reactor core.

No credit is taken for any boron concentration in the BIT, or in any Safety Injection piping downstream of the RWST.

The BIT inlet and outlet isolation valves, (two pairs of motor operated valves, each pair arranged in parallel), are maintained in the open position, as their function to isolate the BIT is not required since implementation of the BIT elimination modification (References 2 through 5).

Maintaining the BIT isolation valves open provides the benefits of eliminating an active safety function and potentially minimizing the time delay in delivery of safety injection flow. The BIT isolation valves may be individually closed for testing (one of each pair at a time) during normal power operation. Only one inlet isolation valve (SI-1852A or B) and one outlet isolation valve (SI-1835A or B) must be open to achieve the emergency core cooling safety function.

However, closing a single BIT isolation MOV presents the potential for loss of the function of the BIT header should a coincident spurious or inadvertent closure of the parallel BIT isolation MOV occur. Spurious or inadvertent mis-positioning of MOVs are considered to be credible single failures. When configured with both parallel BIT inlet or outlet MOVs closed simultaneously, the motor actuators' calculated capabilities lack the opening margin required by the GL 89-10 program. Therefore, in accordance with the IP3 GL 89-10 program, if any of the BIT isolation valves are to be closed in support of maintenance or testing, then the potential for loss of BIT 13 of 215 IPEC00035622 IPEC00035622

IP3 FSAR UPDATE header function via closure of the parallel valve (as caused by any reason including single failure) must be eliminated by administrative means.

A Safety Injection Signal still generates a signal to open the BIT isolation valves. However, based on the limited margin available in the capabilities of the motor actuators of these valves, opening in response to an SI signal would require modification in order to meet the margins required under the GL 89-10 program. Such modifications are unnecessary provided the normal position of the BIT isolation valves is open.

While refueling water has relatively low boron concentration (nominally 2,500 ppm), analyses performed to support implementation of the modification, assumed zero boron concentration in the BIT and associated piping, for conservatism. The Westinghouse "Revised Feasibility Report for BIT Elimination for Indian Point Unit 3," (July 1988) determined that the concentration of boron in the BIT may be reduced to that in the RWST while continuing to meet applicable safety criteria.

The high-head safety injection system is configured as follows:

1. The three available cold leg (one of the four cold leg lines has been permanently isolated by locking closed valve SI-856F on 1-%" Line #754) injection lines on the discharge header containing the retired-in-place BIT are physically connected to the reactor coolant pressure boundary.

2. The three available cold leg (one of the four cold leg lines has been permanently isolated by locking closed valve SI-856A on 2" Line #56) injection lines on the Non-BIT discharge header are physically connected to the accumulator discharge lines upstream of the reactor coolant pressure boundary.

3. The two hot leg injection lines on both discharge headers are physically connected to the reactor coolant pressure boundary.

This configuration was implemented in a modification installed during the 3R13 Refueling Outage to accommodate Stretch Power Uprate to provide for Hot Leg Switchover (HLSO) prior to 6.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />s* following a LBLOCA, and to resolve sump particle and ECCS valve erosion concerns identified in NRC Information Notice 96-27 and 97-76.

Since a small break in the reactor coolant pressure can include a cold leg injection line, safety injection flow capability can be limited by the resulting flow from only five available intact cold leg (Note: one of the original four cold leg lines per header has been isolated by a locked closed valve) injection lines. Depending on the assumed single failure, either two or three safety injection pumps can be operating. To maximize the fraction of safety injection flow delivered to the reactor coolant system with a broken cold leg injection line, the six available cold leg (Note:

one of the original four cold leg lines per header has been isolated by a locked closed valve) injection lines are flow balanced to within an allowable range. The resulting system flow capability is sufficient for the makeup of coolant following a small break that does not immediately depressurize the reactor coolant system to the accumulator discharge pressure.

Credit is not taken for operator action to isolate a broken cold leg injection line.

For large breaks, the Reactor Coolant System would be depressurized and voided of coolant rapidly (about 26 seconds for the largest design break) and a high flow rate is required to quickly recover the exposed fuel rods and limit possible core damage. To achieve this 14 of 215 IPEC00035623 IPEC00035623

IP3 FSAR UPDATE objective, one residual heat removal pump and two safety injection pumps are required to deliver borated water to the cold legs of the reactor coolant loops. Two pumps are available in order to provide for an active component failure. Delivery from these pumps supplements the accumulator discharge. Since the Reactor Coolant System back pressure is relatively low (rapid depressurization for large breaks), a broken injection line would not appreciably change the flows in the other injection lines delivering to the core.

The residual heat removal pumps take suction from the refueling water storage tank.

Because the injection phase of the accident is terminated before the refueling water storage tank is completely emptied, all pipes are kept filled with water before recirculation is initiated.

Water level indication and alarms on the refueling water storage tank give the operator ample warning to terminate the injection phase. Additional level sensors are provided in the containment sump which also give backup indication when injection can be terminated and recirculation initiated.

Recirculation Phases After the injection operation, coolant spilled from the break and water collected from the containment spray is cooled and returned to the Reactor Coolant System by the recirculation system.

Following a Loss-of-Coolant Accident (LOCA), sampling is accomplished as necessary from outside of the Containment via the sampling connection from the recirculating pump discharge.

When the break is large, depressurization occurs due to the large rate of mass and energy loss through the break to Containment. In the event of a large break, the recirculation flow path is within the Containment. The system is arranged so that the recirculation pumps take suction from the recirculation sump in the containment floor and deliver spilled reactor coolant and borated refueling water back to the core through the residual heat exchangers. The system is also arranged to allow either of the residual heat removal pumps to take over the recirculation function. The residual heat removal pumps would only be used if backup capacity to the internal recirculation loop is required. Water is delivered from the Containment to the residual heat removal pumps from a separate sump inside the Containment.

Although the residual heat removal pump is an acceptable alternative for providing core cooling and containment spray flow in lieu of the recirculation pump, there is no single failure that would require its use. The residual heat removal pump(s) would be used only in scenarios beyond the design basis involving multiple active failures. Use of a residual heat removal pump during the long-term recovery phase could be required in the event of ECCS leakage outside Containment.

The motor operated valves in the recirculation suction lines from the containment sump are maintained in the normally closed position at all times, however, they could be opened to allow for residual heat removal pump recirculation operation if that mode was required.

The valves are exercised in accordance with Technical Specification requirements. The valves are operated one at a time and each valve is returned to its normal position before exercising the next one.

No automatic opening features are provided; hence, the probability of a spurious signal to open the valves is nil. The only time these valves are opened is for periodic testing and the 15 of 215 IPEC00035624 IPEC00035624

IP3 FSAR UPDATE procedure ensures that both valves are closed immediately after the test. In addition, the two valves are provided in series to protect against the inadvertent opening of one valve.

The procedure used for periodic testing of these valves ensures that the only water which would be drained from these lines is the small amount trapped between the two valves. This water will discharge to the containment sump. The sump contains two sump pumps which operate on level control and will periodically pump the sump contents to the waste holdup tank during normal plant operation.

For small breaks the depressurization of the Reactor Coolant System is augmented by steam dump and auxiliary feed water addition to the Steam System. For the small breaks in the Reactor Coolant System where recirculated water must be injected against higher pressures for long term core cooling, the system is arranged to deliver the water from the residual heat exchangers to the high-head safety injection pump suction and, by this external recirculation route, to the reactor coolant loops. Thus, if depressurization of the Reactor Coolant System proceeds slowly, the safety injection pumps may be used to augment the flow-pressure capacity of the recirculation pumps in returning the spilled coolant to the reactor.

The recirculation pumps, the residual heat exchangers, piping and valves vital to the function of the recirculation loop are located in a missile-shielded space inside the polar crane support wall on the west side of the reactor primary shield.

There are two recirculation related sumps within the Containment, the recirculation sump and the containment sump. Both sumps collect liquids discharged into the Containment during the injection phase of the design basis accident.

The recirculation sump contains two screens through which the recirculated water must flow before entering the pumps. The first screen consists of a floor grating (1" x 4") which covers the sump on the basement floor. The purpose of the grating is to prevent large particles form entering the sump. The second screen is located in the sump and has the capability to exclude particles greater than 1/8 inch in diameter from the recirculation pump suction. This floor grating has a total surface area of 48.3 fe. Since all recirculated water passes through both screens before entering the pumps, particles in excess of 1/8 inch diameter are precluded from entering these lines. The water velocity through the sump is less than one foot per second.

The containment sump contains two screens for the purpose of preventing particles greater than 1/8 inch diameter from entering the residual heat removal pump suction. The first screen consists of 1" x 4" floor grating with an area of 41.3 fe; the second screen is located in the sump. The water velocity through the sump is less than one foot / second.

The low head external recirculation loop via the containment sump line and the residual heat removal pumps provides backup recirculation capability to the low head internal recirculation loop. The containment sump line is contained within a concentric guard pipe which is connected to the containment liner and terminates within a leak tight compartment. This sump line has two remote motor operated normally closed valves for containment isolation purposes, one of which is within this leak tight compartment.

The high head external recirculation flow path via the high head safety injection pumps is only required for the range of small break sizes for which the Reactor Coolant System pressure remains in excess of the shutoff head of the recirculation pumps (or residual heat removal pumps) at the end of the injection phase or to provide hot leg flow during hot leg recirculation.

16 of 215 IPEC00035625 IPEC00035625

IP3 FSAR UPDATE The external recirculation flow paths within the Primary Auxiliary Building are designed so that external recirculation can be initiated immediately after the accident. Those portions of the Safety Injection System located outside of the Containment which are designed to circulate under post-accident conditions radioactively contaminated water collected in the Containment meet the following requirements:

Shielding to maintain radiation levels within the guidelines set forth in 10 CFR 100 Collection of discharges from pressure relieving devices into closed systems Means to detect and control radioactivity leakage into the environs to the limits consistent with guidelines set forth in 10 CFR 100.

This criterion is met by minimizing leakage from the system. External recirculation loop leakage is discussed in Section 6.2.3.

One pump (either recirculation or residual heat removal) and one residual heat exchanger of the recirculation system provides sufficient cooled recirculated water to keep the core flooded with water by injection through the cold leg connections while simultaneously providing, if required, sufficient containment spray flow to prevent the containment pressure from rising above design limits because of the boiloff from the core. Only one pump and one heat exchanger are required to operate for this capability at the earliest time recirculation is initiated. The design ensures that heat removal from the core and Containment is effective in the event of a pipe or valve body rupture.

Cooling Water The Service Water System (Section 9.6.1) provides cooling water to the component cooling loop, which in turn, cools the residual heat exchangers, all of which are part of the Auxiliary Cooling Systems (Section 9.3). Three conventional service water pumps are available to take suction from the river and discharge to the two component cooling heat exchangers. Three component cooling pumps are available to discharge through their heat exchangers and deliver to the two residual heat exchangers. With the component cooling water system in long term recirculation mode, the following components are required in order to meet core cooling requirements, one residual heat removal pump and heat exchanger, one component cooling water pump, one component cooling water heat exchanger, one service water pump on the nonessential header, and two essential service water pumps on the essential header. All of this equipment with the exception of the residual heat exchangers is located outside Containment.

Containment Building Water Level Monitoring Continuous indication of containment water level during and after an accident is provided by three systems with redundant measuring loops distributed as follows:

Containment Sump (EI. 38' 3"), narrow range, 0' to 10' of water.

Recirculation Sump (EI, 34' 0"), narrow range, 0' to 14' of water.

Containment Building (EI, 46' 0"), wide range, 0' to 8' of water.

Each loop consists of a sensor and a transmitter located inside the containment building, a 17 of 215 IPEC00035626 IPEC00035626

IP3 FSAR UPDATE recorder and power supply at the control room. Refer to Plant Drawing 9321-F-27353 [Formerly Figure No. 6.2-1A].

Change-Over from Injection Phase to Recirculation Phase Assuming that the three high head safety injection pumps, the two residual heat removal pumps, and the two containment spray pumps (Section 6.3) are running at their maximum capacity, the time sequence for the changeover from injection to recirculation in the case of a large rupture beginning form the time of the safety injection signal:

In approximately ten minutes, sufficient water has been delivered to provide the required NPSH to start the recirculation pumps.

In approximately 15 to 20 minutes, (1) one of two low level alarms on the RWST sounds, and the redundant containment recirculation sump level indicators show the sump water level. The alarm serves to alert the operator to start the switchover to the recirculation mode. The redundant containment recirculation sump level indicators provide verification the RWST water has been delivered during the injection phase, in addition to providing consideration to the case of a spurious (i.e., early) RWST low level alarm. The operator would see on the control board that the redundant recirculation sump level indications are at the appropriate points; switch-over to the recirculation phase of safety injection is performed at this time.

With the initiation of the switch sequence (e.g., Switch No.1), only one spray pump will continue in operation. This spray pump will continue to draw from the RWST for approximately 25 minutes to assure that the contents of the spray additive tank have been completely mixed with the spray liquid.

Recirculation pump motors are 2'-2" above the highest water level after addition of the injected water to the spilled coolant.

The entire switchover from injection to recirculation phase is carried out by manually initiating equipment starts/stops and closing a series of switches (each of which carries out several operations) located in the Control Room. At only two pOints in the switchover routine is any reliance on instrumentation necessary:

1) When the low level alarm occurs from the refueling water storage tank low level (LT-920 and/or LlC-921 outside containment), the operator is alerted to begin closing the series of recirculation switches. The low level alarm for each instrument is set to actuate when there is between 10.5 feet and 12.5 feet of water in the tank.

2) After closing switch 4, the operator is required to make a decision whether to close switch 5 or switch 6. The basis for this decision is the flow reading on the flow meters FT-946A, 9468, 946C and 9460. If three or more of these flow meters each indicate greater than zero and the lowest of these readings is at least 360 gpm, .::!:. 10 gpm, the operator will close switch 6; otherwise, the operator will close switch 5.

Analysis indicates that approximately 662 gpm to the core is required to match boil-off at 1398 seconds (the earliest time at which recirculation could be initiated). This includes a 20%

penalty to allow for the effects of hot metal quenching. The flow rates that follow ensure an actual flow of ~662 gpm. Accordingly, a requirement of 360 gpm,.+/- 10 gpm minimum flow rate 18 of 215 IPEC00035627 IPEC00035627

IP3 FSAR UPDATE on the lowest indicating loop has been specified to account for uncertainties in flow measurement and to provide margin.

The decision making process with regard to the flow to the Reactor Coolant System via the low head injection lines is based on readings of the four injection line flowmeters. The rationale for this basis is the following:

For four flow meters each reading greater than zero (i.e., none indicating zero flow):

1) Assume one flow meter fails to an inaccurate high reading, (as a result of a single failure); if the flow rate reads greater than 360 gpm, :!. 10 gpm then this meter is not used as a basis for the 360 gpm, :!. 10 gpm setpoint, but rather, the lowest indicating meter (greater than 0 gpm) is used.

2) Of the four injecting lines, the highest flow line is assumed to be connected to the spilling line; therefore, flow from this line is ineffective.

3) For the three remaining intactiines, their total flow of 751 gpm is delivered to core; if at least 360 gpm, :!.10 gpm is indicated by the lowest line, total flow requirement of 662 gpm is satisfied using low head recirculation.

For one (or two) flow meters each reading greater than zero (i.e., two indicating zero flow):

4} Assume failure of one or two flow meters due to loss of common power supply (i.e.,

single failure). If the flow rate reads greater than 360gpm, :!. 10 gpm then this meter is not used as a basis for the 360 gpm, :!. 10 gpm setpoint, but rather, the lowest indicating meter (greater than 0 gpm) is used.

5) Of the four injecting lines, the highest flow line is assumed to be connected to the spilling line; therefore, flow from this line is ineffective.

6) For the three remaining intact lines, their total flow of 751 gpm is delivered to the core; if at least 360 gpm, :!. 10 gpm is indicated by the lowest line; total flow requirement of 662 gpm is satisfied using low head recirculation.

Thus, the entire recirculation switch over sequence requires only the aforementioned instrumentation. Manual switchover, that is, without use of the "eight-switch sequence," does not require any additional instrumentation; it requires only that the operator close a switch for each and every operation in the switchover sequence.

The control circuitry for the associated valves and pumps is located external to the Containment. The motor operators, associated power cables, and instrumentation inside the Containment have been designed to withstand the LOCA environment as stated in Appendix 6F.

Several motor operated valves operated during the transfer to cold leg or hot leg recirculation are maintained de-energized in their safeguards position during normal power operation in accordance with the Technical Specifications. The subject valves are 8568, 856G, 1810, 882, 744, 842, 843, 883, 1870, 743, 894A, 8948, 894C, and 8940. Each valve (except for the submerged 894C) may be energized during or following the transfer to recirculation at a point in 19 of 215 IPEC00035628 IPEC00035628

IP3 FSAR UPDATE time when spurious or inadvertent mispositioning would not defeat a safety function relied upon to mitigate the consequences of the event.

The manual switch over by the operator which accomplishes the changeover from injection to recirculation is listed below. This switchover takes place when the level indicator or level alarms on the refueling water storage tank indicates that the fluid has been injected. The level indicators in the containment sump will verify that the level is sufficient within the Containment.

The sequence is followed regardless of which power supply is available. The time required to complete the switchover to recirculation is the time for the switch gear to function. All the recirculation switches are grouped together on the safeguard control panel. The service water pumps and component cooling water pumps are located on the auxiliary coolant panel. The component position lights verify when the function of a given switch has been completed.

Should an individual component fail to respond, the operator can take corrective action to secure appropriate response from the back-up component.

The following sequence maintains the loads on the 480V buses within analyzed limits, maintains sufficient core cooling flow during and following the transfer to recirculation and ensures that components are operated within their analyzed limits. While this sequence does not attempt to mirror each procedural step, the major steps listed below must be performed in the sequence described:

1) Terminate safety injection actuation signal and containment spray actuation signal in order that the control logic permits manipulation of the system (at any time following completion of the auto start sequence)

2) Close switch one (remove and isolate unnecessary loads from the diesels).

Trips high head safety injection pump No. 32 if all three are operating (no action if two are operating). Isolates pump No. 32 from the Refueling Water Storage Tank.

Trips spray pump No. 32 if both are operating (no action if only one is operating).

Closes isolation valve at the inoperative spray pump discharge.

3) Close switch three (remove and isolate unnecessary loads from the diesels).

Trips both residual heat removal pumps.

Closes isolation valves at pump suction and discharge headers (the Technical Specifications require the motor operators for these valves to be de-energized).

4) Secure electric auxiliary feedwater pumps prior to closing switch two.

a) If continued Auxiliary Feedwater flow is required, the turbine driven pump is used.

If continued, Auxiliary Feedwater flow is required and the turbine driven pump is unavailable, only one motor driven Auxiliary Feedwater pump may be run.

5) Close switch two (establish cooling flow for Residual Heat Exchangers) 20 of 215 IPEC00035629 IPEC00035629

IP3 FSAR UPDATE a) Starts on one non-essential service water pump (the second or third pump is given a start signal if the first or second pump fails to start).

b) Starts one component cooling water pump (the second or third pump is given a start signal if the first or second pump fails to start).

6) Manually initiate internal recirculation flow.

a) Manually start recirculation Pump A (if Pump A fails to start, use manual start for Pump B; Pump B control switch is adjacent to switch four).

b) Close switch four to open valves on discharge of recirculation pumps. Starting a Recirculation Pump prior to closing switch four minimizes the potential pressure differential across these motor operated valves.

c) Valves SI-HCV-638 and / or SI-HCV-640 are throttled to maintain recirculation flow. For one pump operation, throttling is required to maintain recirculation pump flow within maximum pump flow limits.

7) Check Flow to Reactor Coolant System via the low head injection lines.

a) For the preferred operating mode of omitting switch five and closing switch six (i.e., provides recirculation at low system pressure), the following flow conditions must be verified:

1) With flow in three or more lines greater than zero, the lowest of these flows is at least 360 gpm, +/- 10 gpm.

b) If the above flow conditions are met, the following actions are taken:

1) Direct operators in the field to throttle service ware valves SWN-35-1 and 35-2 to maintain CCW temperature within prescribed limits.

2) Close switch six, which trips operating safety injection pumps.

c) If the above flow conditions are not verified, close switch five and omit switch six (provides recirculation at elevated system pressure).

1) Aligns flow from residual heat exchanger to high head safety injection pumps.

(The motor-operated valves on the outlet of the residual heat exchangers to the suction of the high-head safety injection pumps are opened. The motor-operated valves on the outlets of the residual heat exchangers to the low-head injection lines are closed together with the safety injection pump mini-flow and residual heat removal pump mini-flow).

2) Direct operators in the field to throttle service water valves SWN-35-1 and 35-2 to maintain CCW temperature within prescribed limits.

8) Close switch seven 21 of 215 IPEC00035630 IPEC00035630

IP3 FSAR UPDATE Starts a second non-essential service water pump.

Starts a second component cooling water pump only if the second service water pump successfully started.

Starts a second recirculation pump only if the second component cooling water pump successfully started.

9) Close switch eight (complete the isolation of the safety injection system and containment spray system test lines to the refueling water storage tank).

a) Close the valve on the spray test line.

b) Close the valve in the safety injection pumps suction line from the Refueling Water Storage Tank (control power for this valve is de-energized as required by the Technical Specifications).

If an RHR pump is used for low head recirculation, then valves SI-HVC-638 and/or SI-HCV-640 are throttled to maintain RHR pump flow to the cold legs (and recirculation spray} less than the maximum pump flow limit. Section 9.6.1 describes additional requirements to be met relating to alignment and operation of the service water system at the beginning of the post- LOCA recirculation phase.

Although the listed recirculation switches are manual, each automatically causes the operations listed. An indicating lamp is provided to show the operator when the operations of a given switch have been performed and when he should proceed with the next switching operation. In addition, lamps indicating completion of the individual functions for a given switch are provided.

These lamps are adjacent to the switches. The time required to complete the switch over is just the time for the switch gear to operate. Should an individual component fail to respond, the operator can take corrective action to secure appropriate response from controls within the Control Room. Remote operated valves for the injection phase of the Safety Injection System (Table 6.2-11) which are under manual control (that is, valves which normally are in their ready position and do not receive a safety injection signal) have their positions indicated on a common portion of the control board. At any time during operation when one of these valves is not in the ready position for injection, it is shown visually on the board. Reference is made to Table 6.2-11 which is a listing of the instrumentation readouts on the control board which the operator can monitor during recirculation. In addition, an audible annunciation alerts the operator to the condition.

Hot leg recirculation is initiated after 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> but prior to 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

Location of the Major Components Required for Recirculation The residual heat removal pumps re located in the residual heat removal pump of the Primary Auxiliary Building (EI. 15'). The residual heat exchangers are located on a platform above the basement floor of the Containment Building (EI. 66').

The recirculation pumps are located directly above the recirculation sump in the Containment Building (EI. 46').

22 of 215 IPEC00035631 IPEC00035631

IP3 FSAR UPDATE The component cooling pumps and heat exchangers are located in the Primary Auxiliary Building (EI. 41' and 73', respectively).

The service water pumps are located in the intake structure and the redundant piping to the component cooling heat exchangers is run underground.

Steam Break Protection A large break of a steam system pipe rapidly cools the reactor coolant causing insertion of reactivity into the core and depressurization of the system. Compensation is provided by injection of borated water from the refueling water storage tank (RWST). Redundant isolation valves open upon a safety injection signal, providing a supply of borated water with a boron concentration of 2500 ppm nominally. Even assuming all of the safety injection lines downstream of the RWST, including the BIT, contain unborated water, this is sufficient to terminate the reactor power transient before any clad damage results. The analysis of the steam line rupture accident is presented in Section 14.2.5.

Components All associated components, piping, structures, and power supplies, of the Safety Injection System were designed in accordance with the seismic criteria provided in Section 16.1.1 and were predominately designated as seismic Class 1. Refer to Plant Drawing 9321-F-27353 and -

27503 [Formerly Figures 6.2-1Aand 1B] for indication of the seismic class piping boundaries.

All components inside the Containment are capable of withstanding or are protected from differential pressure which may occur during the rapid pressure rise to 47 psig in 10 seconds.

Emergency core cooling components are austenitic stainless steel, and hence, are quite compatible with the spray solution over the full range of exposure in the post-accident regime.

While this material is subject to crevice corrosion by hot, concentrated caustic, the NaOH additive cannot enter the containment or Emergency Core Cooling Systems without first being diluted and partially neutralized with boric acid to a mild solution. Corrosion tests performed with simulated spray showed negligible attack, both generally and locally, in stressed and unstressed stainless steel at containment and ECCS conditions. These tests are discussed in WCAP-7153(1).

The quality standards of all Safety Injection System components are tabulated in summary form in Table 6.2-12.

Accumulators The accumulators are pressure vessels filled with borated water and pressurized with nitrogen gas. During normal plant operation each accumulator is isolated from the Reactor Coolant System by two check valves in series. Should the Reactor Coolant System pressure fall below the accumulator operating pressure, the check valves open and borated water is forced into the Reactor Coolant System. Mechanical operation of the swing-disc check valves is the only action required to open the injection paths from the accumulators to the core via each cold leg.

Indian Point 3 does not utilize hot leg injection. No timer is involved or provided. The two hot leg connections are provided to allow hot leg recirculation. However, these connections are closed at all times during plant operation.

23 of 215 IPEC00035632 IPEC00035632

IP3 FSAR UPDATE The level of borated water in each accumulator tank is adjusted remotely as required during normal plant operation. During normal plant operation, the fluid level can be reduced by draining through the Sampling System to the Sample Sink in the PAB. The water level can also be reduced by draining to the reactor coolant drain tank or to the VC sump; however, these drain paths degrade the accumulator function by exposing the affected accumulator(s) to non-seismic piping from which it can not be isolated in accordance with design criteria. Draining accumulators to the Reactor Coolant Drain tank or the VC sump may only be performed under the conditions delineated by the plant Technical Specifications. To increase and/or maintain the accumulator water level, refueling water is added using a safety injection pump. Samples of the solution in the tanks are taken at the sampling station for periodic checks of boron concentration.

The accumulators are passive Engineered Safety Features because the gas forces injection; no external source of power or signal transmission is needed to obtain fast-acting, high flow capability when the need arises. On accumulator is attached to each of the cold legs of the Reactor Coolant System.

The design capacity of the accumulators is based on the assumption that flow from one of the accumulators spills onto the containment floor through the ruptured loop. The flow from the three remaining accumulators provides water after the end of blowdown, to reflood the core.

(Section 14.3)

The accumulators are carbon steel, internally clad with stainless steel and designed to ASME Section III, Class C. Connections for remotely draining or filling the fluid space during normal plant operation are provided.

Redundant level and pressure indictors are provided with readouts on the control board. Each indicator is equipped with high and low level alarms.

For the Accumulator Discharge Valves (894 A, B, C, D), the following indications are provided to supervise the administrative procedures and to highlight the existence of an incorrect configuration:

1) Red (open) and Green (closed) position indicating lights at the control switch for each valve. These lights are powered by valve control power and actuated by valve motor operator limit switches.

2) An additional indicating system of lights is used, whereby each valve has a two light sugar cube. The right side of the cube is a WHITE light (which glows PINK if the adjacent RED light is lit) that indicates power applied to the indicating system; the left side of the cube is a RED light to indicate when the respective valve is in its proper position enabling safeguards operation. This grouping highlights a valve not properly lined up. These lights are energized from a separate monitor light supply and the RED light is actuated by a valve motor operator limit switch.

3) In the event a valve is closed for accumulator or valve testing at the time injection is required, a safety injection signal is applied to open the valve (if power is available),

overriding the test closure.

Prior to commercial operation, the AEC required that the electric power to these valves be locked out to prevent a spurious closure. The lockout will be implemented whenever RCS 24 of 215 IPEC00035633 IPEC00035633

IP3 FSAR UPDATE temperature is above 350°F. These valves are closed during plant shutdown conditions to isolate the pressurized accumulators from the depressurized reactor coolant system.

The accumulator design parameters are given in Table 6.2-2.

Boron Injection Tank The Boron Injection Tank (BIT) is "functionally" retired-in-place. That is, it is no longer relied upon to provide concentrated boric acid for injection into the reactor core during emergency core cooling. It does, however, remain a passive component of the Safety Injection System, and as such, it is relied upon for its properties as a pressure vessel. Because the BIT no longer contains concentrated boric acid, the specialized handling requirements associated with that substance, such as heating and recirculation, no longer need to be met. The heaters which reside at the bottom of the BIT have been permanently de-energized. Furthermore, the recirculation flowpath between the BIT and the Boric Acid Storage Tanks has been valved off.

The BIT inlet and outlet isolation valves (two pairs of motor operated valves, each pair arranged in parallel) are maintained in the open position, as their function to isolate the BIT is not required since implementation of the Reference 3 modification. A Safety Injection Signal still generates a signal to open the BIT isolation valves. However, based on the limited margin available in the capabilities of the motor actuators of these valves, opening in response to a SI signal would require modification in order to meet the margins required under the GL 89-10 program. Such modifications are unnecessary provided the normal position of the BIT isolation valves is open.

The BIT remains in the safety injection flowpath, and continues to be relied upon to convey the water contained in it, as well as water from the Refueling Water Storage Tank, in the same manner as a section of piping would. Although the BIT contents are identical to the contents of the RWST, that is, borated water with a nominal boron concentration of 2,500 ppm, no credit is taken by the core response analyses for any boron in the BIT, or the safety injection piping downstream of the RWST, for conservatism.

The design parameters of the BIT are presented in Table 6.2-3.

Refueling Water Storage Tank In addition to its normal duty to supply borated water to the refueling canal for refueling operations, this tank provides borated water to the safety injection pumps, the residual heat removal pumps and the containment spray pumps for the Loss-of-Coolant Accident. During plant operation, it is aligned to these pumps.

The capacity of the refueling water storage tank is based on the requirement for filling the refueling canal. When filled to Technical Specification requirements, approximately 342,200 gallons is available for delivery. One low level alarm is set to actuate at between 10.5 feet and 12.5 feet of water in the tank. This tank capacity and these alarm settings provide an amount of borated water to assure:

1) A sufficient volume of water on the floor to permit the initiation of recirculation (195,800 gal).

25 of 215 IPEC00035634 IPEC00035634

IP3 FSAR UPDATE

2) A volume sufficient to allow switchover to recirculation pumps, containment pressure relief, and sump pH control via containment spray system following a reactor coolant pressure boundary break (66,700 gal).

3) Adequate volume to allow for instrument uncertainties (total 52,100 gallons between the Technical Specifications minimum RWST level of 35.4' and the nominal containment spray shutoff point of 1.5'. Of this volume, 26,100 gallons are eventually added to the Containment, but the remaining 26,000 gallons are considered by the analysis to be unusable.

4) The total RWST volume, when added with accumulator discharge to the reactor coolant system, will assure no return to criticality with the reactor at cold shutdown and no control rods inserted into the core.

The water in the tank is borated to a concentration which assures reactor shutdown by at least 5% cklk when all RCC assemblies are inserted and when the reactor is cooled down for refueling. The maximum boric acid concentration is approximately 1.5 weight percent boric acid. At 32°F the solubility limit of boric acid is 2.2%. Therefore, the concentration of boric acid in the refueling water storage tank is well below the solubility limit at 32°F.

The contents of the Refueling Water Storage Tank are kept above 32°F by a steam heated, austenitic stainless steel pipe coil in the bottom of the tank. Steam is supplied to this coil through a single header from the auxiliary boilers which are used to supply all required auxiliary steam to Indian Point 3.

The passive heating coil and passive single supply header are supplied with steam from any one of five sources. In the remote case of loss of steam to this tank, there would be a time period of at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> available for repair or connection to another steam source before freezing problems would arise, even under the most severe weather conditions. If the electrical heat tracing on the tank discharge line remains operable it is very probable that a freezing problem would not arise.

The steam to the heating coil is automatically flow controlled to maintain a minimum tank water temperature of 35°F. In response to low RWST temperature, steam is admitted by temperature control valve TCV-1116, and the pressure is controlled automatically by pressure control valve PCV-1250 to maintain a nominal 7 psig steam pressure in the coil (see Plant Drawing 9321-F-27273 [Formerly Figure 9.6-16]), During normal and shutdown operations when the tank is filled with borated water, the water pressure outside the heating coil will be approximately 15 psig, thereby preventing leakage of steam out of the coil and subsequent dilution of the borated water.

All outdoor piping connected to the Refueling Water Storage Tank is electrically heat traced.

The failure of any section of heat tracing is annunciated in the Control Room. The power source for the heat tracing can be manually switched between two MCC's, each powered automatically by different emergency diesel generator buses.

The design parameters are presented in Table 6.2-4 Pumps 26 of 215 IPEC00035635 IPEC00035635

IP3 FSAR UPDATE Class I (seismic) pumps in the Emergency Safeguards Systems, their required Net Positive Suction Head (NPSH) at extreme operating conditions, the fluid operating temperature, the NPSH available, the atmospheric pressure assumption, and the elevation of each pump are given in Table 6.2-13.

The Internal Recirculation Pump NPSH data in Table 6.2-13 are given for a single pump operation, which represents the most limiting case for maximizing flow and NPSHR. As shown on the Figure 6.2-4 pump curve, an NPSHR of 12.7 ft is required for a flow of 3530 gpm. These pumps were evaluated to operate under cavitating conditions. The pump vendor has confirmed that reduced levels of NPSH are acceptable (Reference 7), such that the Recirculation Pumps can operate indefinitely with an NPSH value @90% of that required on pump curve (Figure 6.2-4). The effective limit for NPSHR thus becomes 11.4 ft. at an actual flow of 3530 gpm at the pump nozzles. See additional discussion of Recirculation Pump NPSH in Section 6.2.3.

An analysis predicts that, for the large break LOCA, there will be 11.4 ft. of NPSH available, which credits the remainder of the RWST water delivered to the containment prior to the start of recirculation containment spray (References 8, 9, 10). In the case of a small-break LOCA, when elevated RCS pressure would preclude direct low head recirculation, high head recirculation would then be established using the Recirculation Pump(s) to deliver a suction supply to the SI Pumps. During high head recirculation, the Recirculation Pumps operate at lower flow rates and the NPSH requirements are correspondingly lower.

NPSH calculations assume saturated water in the sumps so that no credit is taken for containment pressure exceeding the vapor pressure of the sump water. While this conservative assumption is appropriate at accident initiation, it does not allow any credit for the increase of NPSHA which would result from the gradual cooling of the sump fluid to below saturated conditions.

A review performed pursuant to NRC Generic Letter 85-22 had also established that the actual containment water level would be above the minimum switch over level indicated in Table 6.2-

13. This actual water level would provide sufficient additional NPSH available to overcome the head loss effects of debris which has been postulated to result from the destruction of steam generator thermal insulation by LOCA jet forces.

The three (high head) safety injection pumps for supplying borated water to the Reactor Coolant System are horizontal centrifugal pumps driven by electric motors. Parts of the pump in contact with borated water are stainless steel or equivalent corrosion resistant material. A minimum flow bypass line is provided on each pump discharge to recirculate flow to the refueling water storage tank in the event the pumps are started with the normal flow paths blocked. The bypass line joins a common miniflow line shared by the other pumps. Each safety injection pump is sized at 50% of the capacity required to meet the design criteria outlined in Section 6.2.1. The design parameters are presented in Table 6.2-5, and Figure 6.2-2 gives the performance characteristics of these pumps.

The two residual heat removal (lOW head) pumps of the Auxiliary Coolant System are used to inject borated water at low pressure to the Reactor Coolant System. The two recirculation pumps are used to recirculate fluid from the recirculation sump and send it back to the reactor, the spray headers or to suction of the safety injection pumps. All four of these pumps are of the vertical centrifugal type, driven by electric motors. Parts of the pumps which contact the borated water and sodium hydroxide solution during recirculation are stainless steel or equivalent corrosion resistant material. A minimum flow bypass line is provided on the discharge of the 27 of 215 IPEC00035636 IPEC00035636

IP3 FSAR UPDATE residual heat exchangers to recirculate cooled fluid to the suction of the residual heat removal pumps should these pumps be started with their normal flow paths blocked. Additionally, each residual heat removal pump is provided with a dedicated recirculation line. These recirculation lines prevent either pump from operating at shutoff conditions and, also, during dual-pump operation preclude the stronger residual heat removal pump from dead heading the weaker pump as described in IE Bulletin 88-04. A minimum flow bypass discharging back into the recirculation sump, is provided to protect the recirculation pumps should their normal flow paths be blocked. Figure 6.2-3 and 6.2-4 give the performance characteristics of these pumps. The design parameters are presented in Table 6.2-5.

The safety injection pump bearings are cooled by booster pumps using component cooling water. The booster pumps are directly connected to the injection pump motor shaft. The pump seals were designed to operate at accident conditions without cooling water. Pump data is provided in chapter 9.3.

The recirculation pump motors are enclosed fan cooled. The air is cooled by coils utilizing component cooling water and four auxiliary component cooling pumps located outside the Containment. During recirculation the sump water cools the pump bearings. The four (i.e., two pairs) auxiliary component cooling pumps are started during the injection phase; either pump of a pair is capable of protecting its recirculation pump motor from the containment atmosphere.

The fans are directly connected to the recirculation pump motor shafts. The auxiliary component cooling pumps are a part of the Component Cooling Water System and pump data is provided in Chapter 9. The component cooling water volume constitutes a large heat sink so that the main component cooling pumps are not needed during the injection phase (with loss of offsite power).

Details of the component cooling pumps and service water pumps, which serve the Safety Injection System, are presented in Chapter 9.

The pressure containing parts of the high head safety injection pumps are castings, conforming to ASTM A-296, Grade CA-15. The pressure containing parts of the Residual Heat Removal Pumps and the Recirculation Pumps are castings conforming to ASTM-296, Grade CF-8a (chromium content 21.0 to 22.5) and ASTM-296, Grade CF-8, respectively. Stainless steel forgings were procured per ASTM A-182, Grade F304 or F316, or ASTM A-336, Class F8 or F8M, and stainless plate was constructed to ASTM A-240, type 304 or 316. All bolting material conforms to ASTM A-193. Materials such as weld-deposited Stellite or Colmonoy were used at pOints of close running clearances in the pumps to prevent galling and to assure continued performance ability in high velocity areas subject to erosion.

All pressure containing parts of the pumps were chemically and physically analyzed and the results were checked to ensure conformance with the applicable ASTM specification. In addition, all pressure containing parts of the pump were liquid penetrant inspected in accordance with Appendix VIII of Section VIII of the ASM E Boiler and Pressure Vessel Code.

The acceptance standard for the liquid penetrant test is ANSI B31.1, Code for Pressure Piping, Case N-10.

The pump design was reviewed with special attention to the reliability and maintenance aspects of the working components. Specific areas include evaluation of the shaft seal and bearing design to determine that adequate allowances had been made for shaft deflection and clearances between stationary parts.

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IP3 FSAR UPDATE Where welding of pressure containing parts was necessary, a welding procedure including joint detail was submitted for review and approval by Westinghouse. The procedure included evidence of qualification necessary for compliance with Section IX of the ASM E Boiler and Pressure Vessel Code Welding Qualifications. This requirement also applied to any repair welding performed on pressure containing parts.

The pressure-containing parts of the pump were assembled and hydrostatically tested to 1.5 times the design pressure for 30 minutes.

Each pump was given a complete shop performance test in accordance with Hydraulic Institute Standards. The pumps were run at design flow and head, shut-off head and three additional points to verify performance characteristics. Where NPSH is critical, this value was established at design flow by means of adjusting suction pressure.

Pump Cooling Water Supply Pump Source of Cooling Water

1. Internal Recirculation Pumps Auxiliary Component Cooling Water pumps are used to deliver - 40 gpm to each motor cooler.

2. High Head Safety Injection Pumps Booster pumps connected to the shafts of the SI pumps are designed to circulate - 40 gpm of CCW per pump.

3. Residual Heat Removal Pumps Cooling water for seals is not required when the temperature of the pumped fluid is less than 150 F. This is the case during the 0

injection phase after a LOCA. During the recirculation phase, when the pumped fluid temperature may be more than 150 F, a 0 component cooling pump will be running to supply cooling water to the RHR pump.

4. Containment Spray Pumps This pump pumps fluid with a temperature 0

never in excess of 100 F. Therefore no cooling water is required.

The only period of concern when these pumps experience a lack of cooling water is during the injection phase following a LOCA, since during the recirculation phase and at all other times, the component cooling pumps will be available.

During the injection phase, the only heat removal requirement is for the high head safety injection pumps and for the internal recirculation pumps.

Safety Injection Pumps:

Component cooling water is required for cooling the bearings of these pumps. The heat load is estimated to be 75,000 Btu/hr per pump or a total of 225,000 Btu/hr for 3 pumps.

Internal Recirculation Pumps:

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IP3 FSAR UPDATE Cooling water is required to protect the pump motors from the containment environment during a LOCA. The heat load is estimated to be approximately 150,000 Btu/hr per pump or a total of 300,000 Btu/hr for two pumps.

Since the component cooling pumps do not run during the injection phase, (with loss of offsite power), the water volume of the component cooling system is used as a heat sink. This heat load causes a temperature rise of approximately rF/hour in the component cooling water (no credit is taken for the water volume in the surge tank). With 110 F cooling water at the start of 0

0 the accident, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> are available before the cooling water temperature reaches 150 F; 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> are available before reaching 180 F. 0 Heat Exchangers The two residual heat removal heat exchangers of the Auxiliary Coolant System cool the recirculated sump water. These heat exchangers were sized for the cool-down of the Reactor System. Table 6.2-6 gives the design parameters of the heat exchangers.

The ASME Boiler and Pressure Vessel Code has strict rules regarding the wall thickness of all pressure containing parts, material quality assurance provisions, weld joint design, radiographic and liquid penetrant examination of materials and joints, and hydrostatic testing of the unit as well as requiring final inspection and stamping of the vessel by an ASME Code inspector.

The designs of the heat exchangers also conform to the requirements of TEMA (Tubular Exchanger Manufacturers Association) for Class R heat exchangers. Class R is the most rugged class of TEMA heat exchangers and is intended for units where safety and durability are required under severe service conditions. Items such as: tube spacing, flange design, nozzle location, baffle thickness and spacing, and impingement plate requirements are set forth by TEMA Standards.

In addition to the above, additional design and inspection requirements were imposed to ensure rugged, high quality heat exchangers. The design and inspection requirements included:

confined-type gaskets, main flange studs with two nuts on each end to ensure permanent leak tightness, general construction and mounting brackets suitable for the plant seismic design requirements, tubes and tube sheet capable of withstanding full shell side pressure and temperature with atmospheric pressure on the tube side, ultrasonic inspection in accordance with Paragraph N-324.3 of Section III of the ASME Code of all tubes before bending, penetrant inspection in accordance with Paragraph N-627 of Section III of the ASME Code of all welds and hot or cold formed parts, a hydrostatic test duration of not less than thirty minutes, the witnessing of hydro and penetrant tests by a qualified inspector, a thorough final inspection of the unit for good workmanship and the absence of any gouge marks or other scars that could act as stress concentration pOints, and a review of the radiographs and of the certified chemical and physical test reports for all materials used in the unit.

The residual heat exchangers are conventional vertical shell and U-tube type units. The tubes are seal welded to the tube sheet. The shell connections are flanged to facilitate shell removal for inspection and cleaning of the tube bundle. Each unit has a SA-285 Grade C carbon steel shell, a SA-234 carbon steel shell end cap, SA-213 TP-304 stainless steel tubes, a SA-240 type-304 stainless steel channel, a SA-240 type 304 stainless steel channel cover, and a SA-240 type 304 stainless tube sheet.

Valves 30 of 215 IPEC00035639 IPEC00035639

IP3 FSAR UPDATE All parts of valves used in the Safety Injection System in contact with borated water are austenitic stainless steel or equivalent corrosion resistant material. The motor operators on the injection line isolation valves are capable of rapid operation. All valves required for initiation of safety injection or isolation of the system have remote position indication in the Control Room.

Valving is specified for exceptional tightness, and where possible, instrument valves and packless diaphragm valves are used. All valves, except those which perform a control function, are provided with backseats which are capable of limiting leakage to less than 1.0 cc per hour per inch of stem diameter, assuming no credit for valve packing. Backseats can also be employed to facilitate repacking the valve stem. As a general rule, the plant relies on packing to minimize valve stem leakage. Normally closed globe valves are installed with recirculation flow under the seat to prevent leakage of recirculated water through the valve stem packing. Relief valves are totally enclosed. Control and motor-operated valves which are 2-1/2 in and larger and which are exposed to recirculation flow are provided with double-packed stuffing boxes and stem leakoff connections which are piped to the Waste Disposal System.

The check valves which isolate the Safety Injection System from the Reactor Coolant System are installed immediately adjacent to the reactor coolant piping to reduce the probability of a safety injection line rupture causing a Loss-of-Coolant Accident.

A relief valve is installed in the safety injection pump discharge header discharging to the pressurizer relief tank in order to prevent overpressure in the lines which have a lower design pressure than the Reactor Coolant System. The relief valve is set at the design pressure of the safety injection piping.

The gas relief valves on the accumulators protect them from pressures in excess of the design value.

Motor Operated Gate Valves The pressure containing parts (body, bonnet and discs) of the valves employed in the Safety Injection System were designed per criteria established by the ANSI B16.5 (1955) or MSS SP66 specifications. The materials of construction for these parts were procured per ASTM A 182, F316 or A351, GR-CF8M or CF8. All material in contact with the primary fluid, except the packing, is austenitic stainless steel or equivalent corrosion resisting material. The pressure containing cast components were radiographically inspected as outlined in ASTM E-446 Class 1 or Class 2. The body, bonnet and discs were liquid penetrant inspected in accordance with ASME Boiler and Pressure Vessel Code Section VIII, Appendix VIII. The liquid penetrant acceptable standard was as outlined in ANSI B31.1, Case N-1 o.

When a gasket is employed, the body-to-bonnet joint was designed per ASME Boiler and Pressure Vessel Code Section VIII or ANSI B16.5 with a fully trapped, controlled compression, spiral wound asbestos or suitable material, gasket with provisions for seal welding, or of the pressure seal design with provisions for seal welding. The bOdy-to-bonnet bolting and nut materials were procured per ASTM A 193 and A 194, respectively.

The entire assembled unit was hydrotested as outlined in MSS SP-61 with the exception that the test pressure was maintained for a minimum period of 30 minutes. The seating design of the Darling parallel disc design, the Crane flexible wedge design, or the equivalent. These designs have the feature of releasing the mechanical holding force during the first increment of 31 of 215 IPEC00035640 IPEC00035640

IP3 FSAR UPDATE travel. Thus, the motor operator has to work only against the frictional component of the hydraulic unbalance on the disc and against the packing box friction. The discs are guided throughout the full disc travel to prevent shattering and provide ease of gate movement. The seating surfaces are hard faced (Stellite No.6 or equivalent) to prevent galling and reduce wear.

The stem material is ASTM A276 type 316 condition B or precipitation hardened 17-4 pH stainless, procured and heat treated to Westinghouse specifications. These materials were selected because of their corrosion resistance, high tensile properties, and their resistance to surface scoring by the packing. The valve stuffing box was designed with a lantern ring leakoff connection with a minimum of a full set of packing below the lantern ring and a maximum of one-half of a set of packing above the lantern ring; a full set of packing is defined as a depth of packing equal to 1-1/2 times the stem diameter. The experience with this stuffing box design and the selection of packing and stem materials has been very favorable in both conventional and nuclear power plants.

Valves 744,882,1810, are required to be open during the injection phase of the LOCA and then must be closed for long-term recirculation. There is no time that these valves would be closed during plant power operation. The motors for these valves are normally de-energized with their breakers locked open. In addition, these valves are provided with red/green position indicating lights and monitor lights to highlight valves configuration as described in Section 6.2.2 "Accumulators," items 1 and 2.

Valves 856B and 856G are required to be closed during the injection and cold leg recirculation phases. The motors for these valves are normally de-energized with their breakers locked open and normal indicating lights de-energized. In addition, these valves are interlocked with corresponding cold leg injection line valves on each header to prevent simultaneous opening of all high-head safety lines on each header. The valves are equipped with a position monitor light, via limit switch and separate DC circuit, and an alarm via limit switch.

The motor operator is extremely rugged and is noted throughout the power industry for its reliability. The unit incorporates a "hammer blow" feature that allows the motor to move the valve off its main seat or backseat while allowing the motor to attain its operational speed.

The valve was assembled, hydrostatically tested, seat-leakage tested (fore and back),

operationally tested, cleaned, and packaged per specifications. All manufacturing procedures employed by the valve supplier, such as hard facing, welding, repair welding and testing, were submitted to Westinghouse for approval.

For those valves which function on the safety injection signal, 10 seconds operation or other justified times are typically required. The BIT isolation valves are maintained open, but still receive a safety injection signal to open based on their former application as normally closed valves. An opening stroke time of 11 seconds had been justified for these valves when they were maintained normally closed. However, changing their position from normally closed to normally open has superseded the requirement for a 10 or 11 second opening stroke time. For all other valves in the system, the valve operator completes its cycle from one position to the other typically in 120 seconds.

Valves which must function against system pressure were typically designed such that they function with a pressure drop equal to full system pressure across the valve disc.

Manual Valves 32 of 215 IPEC00035641 IPEC00035641

IP3 FSAR UPDATE The stainless steel manual globe, gate and check valves were designed and built in accordance with the requirements outlined in the motor operated valve description above.

The carbon steel valves were built to conform with ANSI B16.5. The materials of construction of the body, bonnet and disc conform to the requirements of ASTM A105 Grade II, A181 Grad II, or A216 Grade WCB or WCC. The carbon steel valves pass only non-radioactive fluids and were subjected to hydrostatic test as outlined in MSS SP-61 except that the test pressure was maintained for at least 30 minutes. Since the fluid controlled by the carbon steel valves is not radioactive, the double packing and seal weld provisions are not provided.

Accumulator Check Valves The pressure containing parts of this valve assembly were designed in accordance with MSS SP-66. All parts in contact with the operating fluid are of austenitic stainless steel or of equivalent corrosion resistant materials procured to applicable ASTM or WNES specifications.

The cast pressure-containing parts were radiographed in accordance with ASTM E-94 and with the acceptance standard as outlined in ASTM E-446 Class 1 or Class 2. The cast pressure-containing parts, machined surfaces, finished hard facings, and gasket bearing surfaces were liquid penetrant inspected per ASME B&PV Code,Section VIII and the acceptance standard as outlined in ANSI B31.1 Code Case N-10. The final valve was hydrotested per MSS SP-66 except that the test pressure was maintained for at least 30 minutes. The seat leakage was conducted in accordance with the manner prescribed in MSS SP-61.

The valve was designed with a low pressure drop configuration with all operating parts contained within the body, which eliminates those problems associated with packing glands exposed to boric acid. The clapper arm shaft was manufactured from 17-4 pH stainless steel, heat treated to Westinghouse Specifications. The clapper arm shaft bushings were manufactured from Stellite No.6 material. The various working parts were selected for their corrosion resistant, tensile, and bearing properties.

The disc and seat rings were manufactured from a forging. The mating surfaces are hard faced with Stellite No.6 to improve the valve seating life. The disc is permitted to rotate, providing a new seating surface after each valve opening.

The valves are operated in the closed position with a normal differential pressure across the disc of approximately 1700 psi. The valves remain in this position except for testing and safety injection. Since the valve will not be required to normally operate in the open condition, which would subject the valve to impact loads caused by sudden flow reversal, this equipment does not have difficulties performing its required functions.

When the valve is required to function a differential pressure of less than 25 psig will shear any particles that may attempt to prevent the valve from functioning. Although the working parts are exposed to the boric acid solution contained within the reactor coolant loop, a boric acid "freeze up" is not expected with this Iowa concentration.

The experience derived from the check valves employed in the Emergency Injection System of the Carolina-Virginia Tube Reactor in a similar system indicated that the system is reliable and workable. The CVTR Emergency Injection System, maintained at atmospheric conditions, was separated from the main coolant piping by one six inch check valve. Check valve leakage was 33 of 215 IPEC00035642 IPEC00035642

IP3 FSAR UPDATE not a problem. This was further substantiated by the satisfactory experience obtained from operation.

Relief Valves The accumulator relief valves were sized to pass nitrogen gas at a rate in excess of the accumulator gas fill line delivery rate. The relief valves can also pass water in excess of the expected leak rate, but this is not necessary because the time required to fill the gas space gives the operator ample opportunity to correct the situation. For an inleakage rate 15 times the manufacturing test rate, there will be about 1000 days before water will reach the relief valves.

Prior to this, level and pressure alarms would have been actuated.

The safety injection test line relief valves are provided to relieve any pressure above design that might build up in the high head safety injection piping. The valve can pass a nominal 15 gpm (2.25 x 105 cc/hr), which is far in excess of the manufacturing design leak rate of 24 cc/hr.

Leakage Limitations of Valves Valving was specified for exceptional tightness and, where possible, instrument valves, packless diaphragm valves were used.

Normally open valves have backseats which are capable of limiting leakage to less than one cubic centimeter per hour per inch of stem diameter assuming no credit for packing in the valve.

Backseats can also be employed to facilitate repacking the valve stem. As a general rule, the plant relies on packing to minimize valve stem leakage. Normally closed globe valves were installed with recirculation flow under the seat to prevent stem leakage from the more radioactive fluid side of the seat.

Motor operated valves which are exposed to recirculation flow were provided with double-packed stuffing boxes and stem leakoff connections which are piped to the Waste Disposal System.

The specified leakage across the valve disc required to meet the equipment specification and hydrotest requirements is as follows:

Conventional globe - 3 cc/hr/in of nominal pipe size Gate valves - 3 cc/hr/in of nominal pipe size; 10/cc/hr/in for 300 and 150 pound USA Standard Motor operated gate valves - 3 cc/hr/in of nominal pipe size: 10/cC/hr/in for 300 and 150 pound USA Standard Check valves - 3 cc/hr/in of nominal pipe size: 10/cc/hr/in for 300 and 150 pound USA Standard Accumulator check valves - 10 cc/hr/in of nominal pipe size; relief valves are totally enclosed.

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IP3 FSAR UPDATE All Safety Injection System piping in contact with borated water is austenitic stainless steel.

Piping joints are welded except for the flanged connections at the safety injection pumps and recirculation pumps.

The piping beyond the accumulator stop valves was designed for Reactor Coolant System conditions (2485 psig, 650°F). All other piping connected to the accumulator tanks was designed for 700 psig and 400°F.

The safety injection pump and residual heat removal pumps suction piping (210 psig at 300°F) from the refueling water storage meets NPSH requirements of the pumps.

The safety injection high pressure branch lines (1500 psig at 300°F) were designed for high pressure losses to limit the flow rate out of the branch line which may have ruptured at the connection to the reactor coolant loop.

The system design incorporated the ability to isolate the safety injection pumps on separate headers such that full flow from at least one pump is ensured should a branch line break. Two SI pump discharge headers are provided in a configuration which allows 2 of 3 SI Pumps to deliver into either header. The suction flow paths are configured to allow isolation of the 32 SI Pump suction piping from the common suction flow path, with an alternate suction piping alignment dedicated to the 32 SI Pump. The common and alternate suction flow paths are cross-tied via a 0.75" pressure equalization pipe, with two normally closed valves and a normally closed vent valve.

The piping was designed to meet the minimum requirements set forth in (1) the ANSI 831.1 Code (1955) for the Pressure Piping, (2) Nuclear Code Case N-7, (3) ANSI Standards 836.10 and 836.19 and (4) ASTM Standards with supplementary standards plus additional quality control measures.

Minimum wall thickness were determined by the ANSI Code (1955) formula found in the power piping Section 1 of the ANSI Code (1955) for Pressure Piping. This minimum thickness was increased to account for the manufacturer's permissible tolerance of minus 12-1/2 percent on the nominal wall. Purchased pipe and fittings had a specified nominal wall thickness that is no less than the sum of that required for pressure containment, mechanical strength and manufacturing tolerance.

Thermal and seismic piping stress analyses were performed in accordance with ANSI 831.1 code (1967). Special attention was directed to the piping configuration at the pumps with the objective of minimizing pipe imposed loads at the suction and discharge nozzles. Piping is supported to accommodate expansion due to temperature changes during the accident.

Pipe and fittings materials were procured in conformance with all requirements of the ASTM and ANSI specifications. All materials were verified for conformance to specification and documented by certification of compliance to ASTM material requirements. Specifications imposed additional quality control upon the suppliers of pipes and fittings as listed below:

1) Purchased pipe and fittings required the submittal of actual heat chemical and physical test results. Each item or part of a fabrication required identification to an individual test report. Welding materials required the submittal of heat or manufacturers' lot reports showing heat chemical and physical test results.

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IP3 FSAR UPDATE

2) Pipe branch lines 2-1/2 inch and larger between the reactor coolant pipes and the isolation stop valves conform to ASTM A376 and meet the supplementary requirement S6 ultrasonic testing. Fittings conform to the requirements of ASTM A403. Fittings 2-1/2 inch and larger had requirements for UT inspection similar to S6 of A376.

Shop fabrication of piping subassemblies was performed by reputable suppliers in accordance with specifications which defined and governed material procurement, detailed design, shop fabrication, cleaning, inspection, identification, packaging and shipment.

Welds for pipes sized 2-1/2 inch and larger were butt welded. Reducing tees were used where the branch size exceeds ~ of the header size. Branch connections of sizes that are equal to or less than ~ of the header size were of a design that conforms to the ANSI rules for reinforcement set forth in the ANSI B31.1 Code for Pressure Piping. Bosses for branch connections are attached to the header by means of full penetration welds.

All welding was performed by welders and welding procedures qualified in accordance with the ASME Boiler and Pressure Vessel Code Section IX, Welding Qualifications. The Shop Fabricator was required to submit all welding procedures and evidence of qualifications for review and approval prior to release for fabrication. All welding materials used by the Shop Fabricator required prior approval.

All high pressure piping butt welds containing radioactive fluid at greater than 600°F temperature and 600 psig pressure of equivalent were radiographed. The remaining piping butt welds were randomly radiographed. The technique and acceptance standards were those outlined in UW-51 of the ASME B&PV Code Section VIII. In addition, butt welds were liquid penetrant examined in accordance with the procedure of ASME B&PV Code,Section VIII, Appendix VIII and the acceptance standard as defined in the ANSI Nuclear Code Case N-10.

Finished branch welds were liquid penetrant examined on the outside and where size permitted, on the inside root surfaces.

A post-bending solution anneal heat treatment was performed on hot-formed stainless steel pipe bends. Completed bends were then completely cleaned of oxidation from all affected surfaces. The Shop Fabricator was required to submit the bending, heat treatment and cleanup procedures for review and approval prior to release for fabrication.

General cleaning of completed piping subassemblies (inside and outside surfaces) was governed by basic ground rules set forth in the specifications. For example, these specifications prohibited the use of hydrochloric acid and limited the chloride content of service water and demineralized water.

Packaging of the piping subassemblies for shipment was done so as to preclude damage during transit and storage. Openings were closed and sealed with tight-fitting covers to prevent entry of moisture and foreign material. Flange facings and weld end preparations were protected from damage by means of wooden cover plates and securely fastened in position. The packing arrangement proposed by the Shop Fabricator was subject to approval.

Field Run Piping Field running of small diameter piping for essential system including all Engineered Safety Features was not permitted. All seismic Class I and II piping % inch diameter and larger was 36 of 215 IPEC00035645 IPEC00035645

IP3 FSAR UPDATE designed by the architect-engineer. All supports and restraints for that piping were located by the architect-engineer and designed by either the architect-engineer or the subcontractor supplying the pipe support hardware. All seismic Class I stainless steel piping sub-assemblies were prefabricated offsite at one or more subcontractors' pipe fabrication shops. All seismic Class I carbon steel piping subassemblies 2-1/2 inches in diameter and smaller were fabricated in the field.

Certain seismic Class I and II systems comprised of small diameter tubing were field run, for example, the N.S.S.S. Sampling System, which is Class II and utilizes 3/8 diameter tubing.

In instrumentation design, virtually all tubing was field run including tubing for engineered safety related devices. However, the following detailed information was supplied by the architect-engineer where critical design requirements were to be met:

1) Physical location of tubing where separation is required for redundant measurements

2) Detailed design of tubing where thermal expansion of vessels to which tubing is attached requires special expansion loops, etc.

3) Detailed design of typical instrument tubing supports and anchors

4) Detailed design of missile protection of small diameter tubing

5) Detailed design of tubing where proper operation of the instrument is dependent upon adequate slope of lines, etc.

It was found practical to eliminate field running of all seismic Class I and II piping % inch in diameter and larger. However, it was not found practical to limit the use of field running to a greater extent, namely all small diameter seismic Class I and II tubing. Most tubing was erected near the end of the construction phase. At that time, the tubing erection forces had access to more potential support points than were known to the Architect-Engineer. Also, at that time, the construction forces necessarily erected the tubing around objects which would otherwise have been unknown interference during the design phase.

Fabrication of all Class I and II piping, 3/4" diameter and larger, was done by piping subcontractors following orthographic piping drawings prepared by the Architect-Engineer.

These same piping drawings were also used by the A-E to prepare isometric piping drawings which were then used for both the stress analysis by the designer and for installation by the field groups. These isometrics also showed locations of pipe supports and restraints. Any deviation from the drawing required approval by the Architect-Engineer. A copy of the final as-built information was directed to the Architect-Engineer for final design review.

Site Quality Control Procedures were followed embracing: purchasing and receiving inspection to assure that all weld filler materials, field run pipe and fittings met Class I and II quality requirement; joint by joint inspection to assure cleanliness; proper weld fit-up; proper welding and welder certification; and performance of required NDT. All these procedures were in accordance with A-E specifications for installation, ASA 831.1, and Section IX of the ASME Code. Detailed Quality Control records were maintained on a piece by piece basis and were also recorded on approved spool and line isometric drawings to assure that complete Quality Control coverage was obtained.

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IP3 FSAR UPDATE Hydrostatic tests plus hot and/or cold functional tests were performed on completed systems as required and at the appropriate time. No other special quality assurance measures were necessary.

Pump and Valve Motors Motors Outside the Containment Motor electrical insulation systems were supplied in accordance with ANSI, IEEE and NEMA standards and tested as required by such standards.

Temperature rise design selection was such that normal long life is achieved even under accident loading conditions.

Criteria for motors of the Safety Injection System required that under any anticipated mode of operation, the motor name plate 1.15 service factor rating is not to be exceeded. Design and test criteria ensure that motor loading does not exceed the application criteria.

Motors Inside the Containment The SI Recirculation pumps are three stage, vertical pumps driven by 3 phase, 60 cycle, 350 HP motors, and are powered from 480V bus 5A (31) and 480V bus 6A (32). The recirculation pump motors were designed to operate in an ambient condition of saturated steam at 271°F and 47 psig pressure for one day, followed by operation for at least one year at 155°F and 5 psig in a steam atmosphere. The motors are mounted directly to their respective pumps, approximately 2 ft above the highest anticipated water level.

The SI Recirculation Pump motors are provided with thermalastic epoxy insulation and with a heat exchanger. The motors have Class F insulation, temperature rating of 155°C. However, the motor insulation was derated to Class B (130°C) level to provide a safety margin. The operating temperature of the motor insulation is dependent on cooling water temperature rather than the ambient temperature.

The recirculation pump motors are cooled by radiator type coolers using CCW as the cooling medium. Fans are directly connected to the recirculation pump motor shafts. Rotation of the motor rotor and its end fans forces air through the heat exchanger, and air is contained and returned to the ends of the rotor via ducts. A pressure equalizing device permits incident pressure to enter the motor air system so that the bearings are not subject to differential pressures.

The motors are equipped with high temperature grease lubricated ball bearings which would not break down if the bearings were subjected to incident ambient temperatures.

The motors for the valves inside Containment were designed to withstand containment environment conditions following the Loss-of-Coolant Accident so that the valves can perform the required function during the recovery period.

Periodic operation of the motors and tests of the insulation ensure that the motors remain in a reliable operating condition.

38 of 215 IPEC00035647 IPEC0003564 7

IP3 FSAR UPDATE Although the motors which are provided only to drive Engineered Safety Features equipment are normally run only for test, the design loading and temperature rise limits are based on accident conditions. Normal design margins were specified for these motors to make sure the expected lifetime included allowance for the occurrence of accident conditions.

Valve Motor Operators A production line valve motor has been irradiated to a level of 2 x 108 rads using a cobalt-60 irradiation source. The irradiated motor and an identical unirradiated motor have undergone series of reversing tests at room temperature, followed by a series of reversing tests at 275F.

The room temperature test was repeated while vibrating the motors at a frequency of 30 cycles per second. Both motors operated satisfactorily during all of the tests. No significant difference was evident in the comparison of the data for the two units throughout the test period.

Two independent valve operator manufacturers conducted loss of coolant environmental tests on units similar to those used in this plant. Reports of results indicated that all units operated satisfactorily at test conditions more severe than those expected in the loss-of-coolant or steam-break environment for this plant.

In addition, Westinghouse performed environmental tests on a unit similar to that being used in this plant. The results of the Westinghouse tests indicated that the equipment would perform its required function in the post-LOCA environment.

Electrical Supply Details of the normal and emergency power sources for the Safety Injection System are presented in Chapter 8.

Protection Against Dynamic Effects The injection lines penetrate the Containment adjacent to the Primary Auxiliary Building.

For most of the routing, these lines are outside the crane wall, hence, are protected from missiles originating within these areas. Each line penetrates the crane wall near the injection point to the reactor coolant pipe. In this manner, maximum separation, hence, protection is provided in the coolant loop area.

In the event of a Loss-of-Coolant Accident, all piping systems required to function are designed to remain within acceptable stress limits. The stresses due to dead weight, pressure, operational or design basis earthquake, and maximum motions of the Reactor Coolant Loop imposed on the attached Safety Injection Piping were evaluated in accordance with the stress limits in Section 16.1. The inclusion of the stresses in the injection lines required to function due to movements of the Reactor Coolant Loop assures that these lines maintain their integrity during a Loss-of-Coolant Accident.

All piping supports were designed for the loads imposed by the supported system. The rated loads of allowable stress limits for standard manufactured support components are in accordance with requirements of MSS-SP-58-1967. For non-standard supports designed by analysis, the requirements of AISC-1969 were followed. Where support integrity is dependent on reinforced concrete anchorage, the design was in accordance with the Requirements of ACI-318-63.

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IP3 FSAR UPDATE These standards provide minimum requirements on materials, design and fabrication with ample safety margins for both dead and dynamic loads over the life of the equipment.

Specifically, these standards required that:

1) All materials used be in accordance with ASTM specifications which establish quality levels for the manufacturing process, minimum strength properties, and for test requirements which ensure compliance with the specifications

2) There be proper qualification of welding processes and welders for each class of material welded and for types and positions of welds

3) Maximum allowable stress values be established which provide an ample safety margin on both yield strength and ultimate strength.

6.2.3 Design Evaluation Range of Core Protection The measure of effectiveness of the Safety Injection System is the ability of the pumps and accumulators to keep the core flooded or to reflood the core rapidly where the core has been uncovered for postulated large area ruptures. The result of this performance is to sufficiently limit any increase in clad temperature below a value where emergency core cooling objectives are met. (See Section 6.2.1)

With minimum onsite emergency power available (two-of-three diesel generators), the emergency core cooling equipment consists of two out of three safety injection pumps, one or two out of two residual heat pumps, and three out of four accumulators for a cold leg break and four accumulators for a hot leg break. With these systems, the calculated maximum fuel cladding temperature is limited to a temperature less than that which meets the emergency core cooling design objectives for all break sizes up to and including the double-ended severance of the reactor coolant pipe. (See Section 14.3)

For large area ruptures analyzed (see Section 14.3) the clad temperatures are turned around by the accumulator injection. The active pumping components serve only to complete the refill started by the accumulators. Either two safety injection pumps or one residual heat removal pump provides sufficient addition of water to continue the reduction of clad temperature initially caused by the accumulator.

System Response To provide protection for large area ruptures in the Reactor Coolant System, the Safety Injection System must respond to rapidly reflood the core following the depressurization and core voiding that is characteristic of large area ruptures. The accumulators act to perform the rapid reflooding function with no dependence on the normal or emergency power sources, and also with no dependence on the receipt of an actuation signal.

Operation of this system with three of the four available accumulators delivering their contents to the reactor vessel (one accumulator spilling through the break) prevents fuel clad melting and limits metal-water reaction to an insignificant amount (less than 1%).

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IP3 FSAR UPDATE The function of the safety injection or residual heat removal pumps is to complete the refill of the vessel and ultimately return the core to a sub-cooled state. The flow from either two safety injection pumps or one residual heat removal pump is sufficient to complete the refill with no loss of level in the core.

The design features applied to the Residual Heat Removal System (RHRS) Valves 730 and 731, that isolate it from the Reactor Coolant System provide a diverse combination of control interlock and mechanical limitations preventing improper opening of these valves and also pressure relief capacity capable of limiting pressure if the valves are not closed upon startup of the plant. These features are:

1) That the valves that are separately interlocked with independent pressure control signals to prevent their being opened whenever the Reactor Coolant System pressure is greater than a designated setpoint (which is below the RHRS design pressure).

The pressure interlock was not specifically designed to meet the requirements of IEEE Standard 279-1971. However, each valve, its associated pressure channel and related circuitry are powered from separate instrument buses, and wiring separation is provided to preclude any single failure from rendering both of the valves' control circuits inoperable. Each of the pressure channels is provided with separate Control Room indication to show channel operability.

A separate pressure interlock is provided for each of the two Valves Nos. 730 and 731. Each pressure interlock prevents its valve from being opened when the Reactor Coolant System pressure is greater than a designated open permissive setpoint and also automatically closes the valve whenever the Reactor Coolant System pressure is above a designated auto-close setpoint. These setpoints are below the design pressure of the RHRS.

While the automatic closure interlock for MOV-730 and -731 will prevent over-pressurizing the RHR system piping during an RCS pressure increase transient, this interlock will isolate the suction source of the operating RHR pump(s), potentially causing pump failure. In order to prevent inadvertent isolation of the RHR pump suction, this auto-closure interlock may be defeated by de-energizing the motor operators to MOV-730 and -731. Prior to de-energizing these MOV's Reactor Coolant System Tave must be below 200°F, depressurized and vented through a minimum equivalent opening of two (2) square inches.

2) That the Reactor Coolant System pressure interlocks meet single failure criteria.

3) That the motors are qualified in accordance with IEEE 323-1974, IEEE 344-1975, IEEE 382-1972 for increased reliability and operability in the normal and accident containment environment.

The Residual Heat Removal System was designed for a pressure of 600 psig and 400°F and was hydrostatically tested at a pressure of 900 psig prior to initial operation. Insofar as the piping itself is concerned, the piping code (USAS 831.1) allows a rating of 700 psig at 400°F for schedule 40 stainless steel pipe. Thus the piping system, as presently designed, incorporates a considerable margin in that it is rated at a pressure-temperature condition which is less than that allowed by Code. It 41 of 215 IPEC00035650 IPEC00035650